Sample records for thermal neutron detector from the National Library of Energy Beta (NLEBeta)

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Two-dimensional position sensitive detectors are indispensable in neutron diffraction experiments for determination of molecular and crystal structures in biology, solid-state physics and polymer chemistry. Some performance characteristics of these detectors are elementary and obvious, such as the position resolution, number of resolution elements, neutron detection efficiency, counting rate and sensitivity to gamma-ray background. High performance detectors are distinguished by more subtle characteristics such as the stability of the response (efficiency) versus position, stability of the recorded neutron positions, dynamic range, blooming or halo effects. While relatively few of them are needed around the world, these high performance devices are sophisticated and fairly complex; their development requires very specialized efforts. In this context, we describe here a program of detector development, based on {sup 3}He filled proportional chambers, which has been underway for some years at Brookhaven. Fundamental approaches and practical considerations are outlined that have resulted in a series of high performance detectors with the best known position resolution, position stability, uniformity of reliability over time of this type.

Pillar detector is an innovative solid state device structure that leverages advanced semiconductor fabrication technology to produce a device for thermalneutron detection. State-of-the-art thermalneutrondetectors have shortcomings in achieving simultaneously high efficiency, low operating voltage while maintaining adequate fieldability performance. By using a 3-dimensional silicon PIN diode pillar array filled with isotopic boron 10, ({sup 10}B) a high efficiency device is theoretically possible. The fabricated pillar structures reported in this work are composed of 2 {micro}m diameter silicon pillars with a 4 {micro}m pitch and pillar heights of 6 and 12 {micro}m. The pillar detector with a 12 {micro}m height achieved a thermalneutron detection efficiency of 7.3% at 2V.

Solid-state thermalneutrondetectors are desired to replace {sup 3}He tube based technology for the detection of special nuclear materials. {sup 3}He tubes have some issues with stability, sensitivity to microphonics and very recently, a shortage of {sup 3}He. There are numerous solid-state approaches being investigated that utilize various architectures and material combinations. By using the combination of high-aspect-ratio silicon PIN pillars, which are 2 {micro}m wide with a 2 {micro}m separation, arranged in a square matrix, and surrounded by {sup 10}B, the neutron converter material, a high efficiency thermalneutrondetector is possible. Besides intrinsic neutron detection efficiency, neutron to gamma discrimination is an important figure of merit for unambiguous signal identification. In this work, theoretical calculations and experimental measurements are conducted to determine the effect of structure design of pillar structured thermalneutrondetectors including: intrinsic layer thickness, pillar height, substrate doping and incident gamma energy on neutron to gamma discrimination.

According to the present invention, a system for measuring a thermalneutron emission from a neutron source, has a reflector/moderator proximate the neutron source that reflects and moderates neutrons from the neutron source. The reflector/moderator further directs thermalneutrons toward an unmoderated thermalneutrondetector.

Purpose: A rhodium self-powered neutrondetector (Rh SPND) has been specifically developed by the Comision Nacional de Energia Atomica (CNEA) of Argentina to measure locally and in real time thermalneutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. Methods: The first calibration of the detector was done with the well-thermalizedneutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermalneutron sensitivities were determined from these measurements. Thermalneutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Results: Local mixed-field thermalneutron sensitivities and global thermal and mixed-field thermalneutron sensitivities derived from measurements performed at the RA-6 were compared and no significant differences were found. Global RA-6-based thermalneutron sensitivity showed agreement with pure thermalneutron sensitivity measurements performed in the RA-3 spectrum. Additionally, the detector response proved nearly unchanged by differences in neutron spectra from real (RA-6 BNCT beam) and ideal (considered for calibration calculations at RA-3) neutron source descriptions. The results confirm that the special design of the Rh SPND can be considered as having a pure thermal response for neutron spectra with epithermal-to-thermal flux ratios up to 12%. In addition, the linear response of the detector to thermal flux allows the use of a mixed-field thermalneutron sensitivity of 1.95 {+-} 0.05 x 10{sup -21} A n{sup -1}{center_dot}cm{sup 2}{center_dot}s. This sensitivity can be used in spectra with up to 21% epithermal-to-thermal flux ratio without significant error due to epithermal neutron and gamma induced effects. The values of the measured fluxes in clinical applications had discrepancies with calculated results that were in the range of -25% to +30%, which shows the importance of a local on-line independent measurement as part of a treatment planning quality control system. Conclusions: The usefulness of the CNEA Rh SPND for the on-line local measurement of thermalneutron flux on BNCT patients has been demonstrated based on an appropriate neutron spectra calibration and clinical applications.

In this work, we investigate the optimal thickness of a semiconductor diode for thin-film solid state thermalneutrondetectors. We evaluate several diode materials, Si, CdTe, GaAs, C (diamond), and ZnO, and two neutron converter materials, {sup 10}B and {sup 6}LiF. Investigating a coplanar diode/converter geometry, we determine the minimum semiconductor thickness needed to achieve maximum neutron detection efficiency. By keeping the semiconductor thickness to a minimum, gamma rejection is kept as high as possible. In this way, we optimize detector performance for different thin-film semiconductor materials.

A detector of bursts of neutrons from a deuterium-deuteron reaction includes a quantity of arsenic adjacent a gamma detector such as a scintillator and photomultiplier tube. The arsenic is activated by the 2.5-MeV neutrons to release gamma radiation which is detected to give a quantitative representation of detected neutrons.

Room temperature operating solid state hand held neutrondetectors integrate one or more relatively thin layers of a high neutron interaction cross-section element or materials with semiconductor detectors. The high neutron interaction cross-section element (e.g., Gd, B or Li) or materials comprising at least one high neutron interaction cross-section element can be in the form of unstructured layers or micro- or nano-structured arrays. Such architecture provides high efficiency neutrondetector devices by capturing substantially more carriers produced from high energy .alpha.-particles or .gamma.-photons generated by neutron interaction.

A neutrondetector of very high temporal resolution is described. It may be used to measure distributions of neutrons produced by fusion reactions that persist for times as short as about 50 picoseconds.

Objects of various shapes, with some appreciable hydrogen content, were exposed to fast neutrons from a pulsed D-T generator, resulting in a partially-moderated spectrum of backscattered neutrons. The thermal component of the backscatter was used to form images of the objects by means of a coded aperture thermalneutron imaging system. Timing signals from the neutron generator were used to gate the detection system so as to record only events consistent with thermalneutrons traveling the distance between the target and the detector. It was shown that this time-of-flight method provided a significant improvement in image contrast compared to counting all events detected by the position-sensitive {sup 3}He proportional chamber used in the imager. The technique may have application in the detection and shape-determination of land mines, particularly non-metallic types.

We have developed wavelength-Shifting-fiber Scintillator Detector (SSD) with 0.3 m2 area per module. Each module has 154 x 7 pixels and a 5 mm x 50 mm pixel size. Our goal is to design a large area neutrondetector offering higher detection efficiency and higher count-rate capability for Time-Of-Flight (TOF) neutron diffraction in Spallation Neutron Source (SNS). A ZnS/6LiF scintillator combined with a novel fiber encoding scheme was used to record the neutron events. A channel read-out-card (CROC) based digital-signal processing electronics and position-determination algorithm was applied for neutron imaging. Neutron-gamma discrimination was carried out using pulse-shape discrimination (PSD). A sandwich flat-scintillator detector can have detection efficiency close to He-3 tubes (about 10 atm). A single layer flat-scintillator detector has count rate capability of 6,500 cps/cm2, which is acceptable for powder diffractometers at SNS.

A STUDY OF THE SAL NEUTRONDETECTOR EFFICIENCY USING PHOTODISINTEGRATION OF THE DEUTERON A Thesis for the Degree of Master of Science in the Department of Physics and Engineering Physics University of Physics and Engineering Physics University of Saskatchewan Saskatoon, Saskatchewan S7N 0W0 i #12;Abstract

The detection efficiency, or sensitivity, of a neutrondetector material such as of Si, SiC, amorphous Si, GaAs, or diamond is substantially increased by forming one or more cavities, or holes, in its surface. A neutron reactive material such as of elemental, or any compound of, .sup.10 B, .sup.6 Li, .sup.6 LiF, U, or Gd is deposited on the surface of the detector material so as to be disposed within the cavities therein. The portions of the neutron reactive material extending into the detector material substantially increase the probability of an energetic neutron reaction product in the form of a charged particle being directed into and detected by the neutrondetector material.

Simultaneous detection of gamma rays and neutrons emanating from an unknown source has been of special significance and importance to consequence management and first responders since the earliest days of the program. Bechtel Nevada scientists have worked with 6LiI(Eu) crystals and 6Li glass to develop field-operable dual sensors that detect gamma rays and neutrons simultaneously. The prototype 6LiI(Eu) counter, which has been built and tested, is designed to collect data for periods of one second to more than eight hours. The collection time is controlled by thumbwheel switches. A fourpole, high pass filter at 90 KHz reduces microphonic noise from shock or vibration. 6LiI(Eu) crystals generate completely separable gamma-ray and thermalneutron responses. The 6LiI(Eu) rate meter consists of a single crystal 3.8 x 3.8 cm (1.5 x 1.5 in) with a 2.54-cm-(1-in-) thick, annular, high-density, polyethylene ring around the cylinder. Special features are (1) thermal and epithermal neutron detection (0.025eV to 250keV) and (2) typical gamma resolution of 8% at 661.6 keV. Monte Carlo N-Particle calculations for characteristics of gamma spectral behavior, neutron attenuation length, relative neutron and gamma detection efficiency are reported.

A pulsed neutrondetector and system for detecting low intensity fast neutron pulses has a body of beryllium adjacent a body of hydrogenous material the latter of which acts as a beta particle detector, scintillator, and moderator. The fast neutrons (defined as having En>1.5 MeV) react in the beryllium and the hydrogenous material to produce larger numbers of slow neutrons than would be generated in the beryllium itself and which in the beryllium generate hellium-6 which decays and yields beta particles. The beta particles reach the hydrogenous material which scintillates to yield light of intensity related to the number of fast neutrons. A photomultiplier adjacent the hydrogenous material (scintillator) senses the light emission from the scintillator. Utilization means, such as a summing device, sums the pulses from the photo-multiplier for monitoring or other purposes.

The invention comprises a neutrondetector (50) of very high temporal resolution that is particularly well suited for measuring the fusion reaction neutrons produced by laser-driven inertial confinement fusion targets. The detector comprises a biased two-conductor traveling-wave transmission line (54, 56, 58, 68) having a uranium cathode (60) and a phosphor anode (62) as respective parts of the two conductors. A charge line and Auston switch assembly (70, 72, 74) launch an electric field pulse along the transmission line. Neutrons striking the uranium cathode at a location where the field pulse is passing, are enabled to strike the phosphor anode and produce light that is recorded on photographic film (64). The transmission line may be variously configured to achieve specific experimental goals.

Radiation portal monitors used for interdiction of illicit materials at borders include highly sensitive neutron detection systems. The main reason for having neutron detection capability is to detect fission neutrons from plutonium. The currently deployed radiation portal monitors (RPMs) from Ludlum and Science Applications International Corporation (SAIC) use neutrondetectors based upon 3He-filled gas proportional counters, which are the most common large neutrondetector. There is a declining supply of 3He in the world, and thus, methods to reduce the use of this gas in RPMs with minimal changes to the current system designs and sensitivity to cargo-borne neutrons are being investigated. Reported here are the results of tests of the 6Li/ZnS(Ag)-coated non-scintillating plastic fibers option. This testing measured the required performance for neutron detection efficiency and gamma ray rejection capabilities of a system manufactured by Innovative American Technology (IAT).

High-efficiency neutrondetector substrate assemblies comprising a first conductive substrate, wherein a first side of the substrate is in direct contact with a first layer of a powder material comprising .sup.10boron, .sup.10boron carbide or combinations thereof, and wherein a conductive material is in proximity to the first layer of powder material; and processes of making said neutrondetector substrate assemblies.

Sample records for thermal neutron detector from the National Library of Energy Beta (NLEBeta)

Note: This page contains sample records for the topic "thermal neutron detector" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
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A device for detection of neutrons comprises: as an active neutron sensing element, a conductive organic polymer having an electrical conductivity and a cross-section for said neutrons whereby a detectable change in said conductivity is caused by impingement of said neutrons on the conductive organic polymer which is responsive to a property of said polymer which is altered by impingement of said neutrons on the polymer; and means for associating a change in said alterable property with the presence of neutrons at the location of said device.

This work demonstrated the feasibility and limitations of semiconducting {pi}-conjugated organic polymers for fast neutron detection via n-p elastic scattering. Charge collection in conjugated polymers in the family of substituted poly(p-phenylene vinylene)s (PPV) was evaluated using band-edge laser and proton beam ionization. These semiconducting materials can have high H/C ratio, wide bandgap, high resistivity and high dielectric strength, allowing high field operation with low leakage current and capacitance noise. The materials can also be solution cast, allowing possible low-cost radiation detector fabrication and scale-up. However, improvements in charge collection efficiency are necessary in order to achieve single particle detection with a reasonable sensitivity. The work examined processing variables, additives and environmental effects. Proton beam exposure was used to verify particle sensitivity and radiation hardness to a total exposure of approximately 1 MRAD. Conductivity exhibited sensitivity to temperature and humidity. The effects of molecular ordering were investigated in stretched films, and FTIR was used to quantify the order in films using the Hermans orientation function. The photoconductive response approximately doubled for stretch-aligned films with the stretch direction parallel to the electric field direction, when compared to as-cast films. The response was decreased when the stretch direction was orthogonal to the electric field. Stretch-aligned films also exhibited a significant sensitivity to the polarization of the laser excitation, whereas drop-cast films showed none, indicating improved mobility along the backbone, but poor {pi}-overlap in the orthogonal direction. Drop-cast composites of PPV with substituted fullerenes showed approximately a two order of magnitude increase in photoresponse, nearly independent of nanoparticle concentration. Interestingly, stretch-aligned composite films showed a substantial decrease in photoresponse with increasing stretch ratio. Other additives examined, including small molecules and cosolvents, did not cause any significant increase in photoresponse. Finally, we discovered an inverse-geometric particle track effect wherein increased track lengths created by tilting the detector off normal incidence resulted in decreased signal collection. This is interpreted as a trap-filling effect, leading to increased carrier mobility along the particle track direction. Estimated collection efficiency along the track direction was near 20 electrons/micron of track length, sufficient for particle counting in 50 micron thick films.

The need for monitoring weapons grade Pu in nuclear facilities worldwide was addressed with four radiation detector technologies being developed at Y-12 and ORNL. This paper describes experimental results of 4 Oak Ridge Sensors for Enhancing Nuclear Safeguards (ORSENS) neutrondetector technologies and includes the potential application, cost, and advantages for each. These are a {sup 6}LiF- ZnS(Ag) thermalneutron scintillator coupled to a wavelength-shifting optical fiber, a CdWO{sub 4} based scintillating thermalneutrondetector, a rhodium silicon thermalneutrondetector, and a proton- recoil fast neutrondetector.

Various large-scale neutron sources already build or to be constructed, are important for materials research and life science research. For all these neutron sources, neutrondetectors are very important aspect. However, there is a lack of a high-performance and low-cost neutron beam monitor that provides time and temporal resolution. The objective of this SBIR Phase I research, collaboratively performed by Midwest Optoelectronics, LLC (MWOE), the University of Toledo (UT) and Oak Ridge National Laboratory (ORNL), is to demonstrate the feasibility for amorphous silicon based neutron beam monitors that are pixilated, reliable, durable, fully packaged, and fabricated with high yield using low-cost method. During the Phase I effort, work as been focused in the following areas: 1) Deposition of high quality, low-defect-density, low-stress a-Si films using very high frequency plasma enhanced chemical vapor deposition (VHF PECVD) at high deposition rate and with low device shunting; 2) Fabrication of Si/SiO2/metal/p/i/n/metal/n/i/p/metal/SiO2/ device for the detection of alpha particles which are daughter particles of neutrons through appropriate nuclear reactions; and 3) Testing of various devices fabricated for alpha and neutron detection; As the main results: · High quality, low-defect-density, low-stress a-Si films have been successfully deposited using VHF PECVD on various low-cost substrates; · Various single-junction and double junction detector devices have been fabricated; · The detector devices fabricated have been systematically tested and analyzed. · Some of the fabricated devices are found to successfully detect alpha particles. Further research is required to bring this Phase I work beyond the feasibility demonstration toward the final prototype devices. The success of this project will lead to a high-performance, low-cost, X-Y pixilated neutron beam monitor that could be used in all of the neutron facilities worldwide. In addition, the technologies developed here could be used to develop X-ray and neutron monitors that could be used in the future for security checks at the airports and other critical facilities. The project would lead to devices that could significantly enhance the performance of multi-billion dollar neutron source facilities in the US and bring our nation to the forefront of neutron beam sciences and technologies which have enormous impact to materials, life science and military research and applications.

The disclosure is directed to an apparatus and method for determining the content and distribution of a thermalneutron absorbing material within an object. Neutrons having an energy higher than thermalneutrons are generated and thermalized. The thermalneutrons are detected and counted. The object is placed between the neutron generator and the neutrondetector. The reduction in the neutron flux corresponds to the amount of thermalneutron absorbing material in the object. The object is advanced past the neutron generator and neutrondetector to obtain neutron flux data for each segment of the object. The object may comprise a space reactor heat pipe and the thermalneutron absorbing material may comprise lithium.

Radiation portal monitors used for interdiction of illicit materials at borders include highly sensitive neutron detection systems. The main reason for having neutron detection capability is to detect fission neutrons from plutonium. The currently deployed radiation portal monitors (RPMs) from Ludlum and Science Applications International Corporation (SAIC) use neutrondetectors based upon 3He-filled gas proportional counters, which are the most common large neutrondetector. There is a declining supply of 3He in the world; thus, methods to reduce the use of this gas in RPMs with minimal changes to the current system designs and detection capabilities are being investigated. Reported here are the results of tests of the efficiency of BF3 tubes at a pressure of 800 torr. These measurements were made partially to validate models of the RPM system that have been modified to simulate the performance of BF3-filled tubes. While BF3 could be a potential replacement for 3He, there are limitations to its use in deployed systems.

A novel thermal reactor was designed in which a first reflector formed from a high atomic weight, nonmoderating material is disposed immediately adjacent to the reactor core. A second reflector composed of a moderating material is disposed outwardly of the first reflector. The advantage of this novel reflector arrangement is that the first reflector provides a high slow neutron flux in the second reflector, where irradiation experiments may be conducted with a small effect on reactor reactivity.

DetectorsDetectorsDetectors The detector design group, led by Yacouba Diawara is responsible for supporting the design of HFIR and SNS instruments by developing the necessary infrastructure and acquiring detector components that will be used to complete the functionality of the instruments. The group's mission also includes supporting detector research and development (R&D) for the various instruments and their different needs. The support effort for instrument design entails monitoring detector development worldwide as neutron facilities around the globe are getting upgraded and adopting the newest technologies. Detector group technician Ted Visscher inspects a parahedreal lens on an Anger camera Detector group technician Ted Visscher inspects a parahedreal lens on an

As part of the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program, a joint Idaho State University (ISU) / French Alternative Energies and Atomic Energy Commission (CEA) / Idaho National Laboratory (INL) project was initiated in FY-10 to investigate the feasibility of using neutron sensors to provide online measurements of the neutron flux and fission reaction rate in the ATR Critical Facility (ATRC). A second objective was to provide initial neutron spectrum and flux distribution information for physics modeling and code validation using neutron activation based techniques in ATRC as well as ATR during depressurized operations. Detailed activation spectrometry measurements were made in the flux traps and in selected fuel elements, along with standard fission rate distribution measurements at selected core locations. These measurements provide additional calibration data for the real-time sensors of interest as well as provide benchmark neutronics data that will be useful for the ATR Life Extension Program (LEP) Computational Methods and V&V Upgrade project. As part of this effort, techniques developed by Prof. George Imel will be applied by Idaho State University (ISU) for assessing the performance of various flux detectors to develop detailed procedures for initial and follow-on calibrations of these sensors. In addition to comparing data obtained from each type of detector, calculations will be performed to assess the performance of and reduce uncertainties in flux detection sensors and compare data obtained from these sensors with existing integral methods employed at the ATRC. The neutrondetectors required for this project were provided to team participants at no cost. Activation detectors (foils and wires) from an existing, well-characterized INL inventory were employed. Furthermore, as part of an on-going ATR NSUF international cooperation, the CEA sent INL three miniature fission chambers (one for detecting fast flux and two for detecting thermal flux) with associated electronics for assessment. In addition, Prof. Imel, ISU, has access to an inventory of Self-Powered NeutronDetectors (SPNDs) with a range of response times as well as Back-to-Back (BTB) fission chambers from prior research he conducted at the Transient REActor Test Facility (TREAT) facility and Neutron RADiography (NRAD) reactors. Finally, SPNDs from the National Atomic Energy Commission of Argentina (CNEA) were provided in connection with the INL effort to upgrade ATR computational methods and V&V protocols that are underway as part of the ATR LEP. Work during fiscal year 2010 (FY10) focussed on design and construction of Experiment Guide Tubes (EGTs) for positioning the flux detectors in the ATRC N-16 locations as well as obtaining ATRC staff concurrence for the detector evaluations. Initial evaluations with CEA researchers were also started in FY10 but were cut short due to reactor reliability issues. Reactor availability issues caused experimental work to be delayed during FY11/12. In FY13, work resumed; and evaluations were completed. The objective of this "Quick Look" report is to summarize experimental activities performed from April 4, 2013 through May 16, 2013.

An improved neutron responsive self-powered radiation detector is disclosed in which the neutron absorptive central emitter has a substantially neutron transmissive conductor collector sheath spaced about the emitter and the space between the emitter and collector sheath is evacuated.

Detectors âș R & D 100 Award Detectors âș R & D 100 Award ORNL team wins R&D 100 award for wavelength-shifting scintillator detectorNeutron facilities, national security monitoring will benefit from high-accuracy detector June 2012, Written by Agatha Bardoel A team of eight scientists and technicians in the Neutron Sciences Directorate has won a prestigious R&D 100 Award from R&D Magazine for developing a highly efficient new detector system that helps take pressure off dwindling worldwide supplies of 3He as an active neutron converter. Members of the team receiving an R&D 100 Award for the wavelength-shifting scintillator detector Members of the team receiving an R&D 100 Award for the wavelength-shifting scintillator detector are shown with their invention. They are (from left)

A first study of neutron tagging is conducted in Super--Kamiokande, a 50,000-ton water Cherenkov detector. The tagging efficiencies of thermalneutrons are evaluated in a 0.2 % GdCl$_{3}$-water solution and pure water. They are determined to be, respectively, 66.7 % for events above 3 MeV and 20 % with corresponding background probabilities of 2 $\\times$ 10$^{-4}$ and 3 $\\times$ 10$^{-2}$. This newly developed technique may enable water Cherenkov detectors to identify $\\bar \

A first study of neutron tagging is conducted in Super--Kamiokande, a 50,000-ton water Cherenkov detector. The tagging efficiencies of thermalneutrons are evaluated in a 0.2 % GdCl$_{3}$-water solution and pure water. They are determined to be, respectively, 66.7 % for events above 3 MeV and 20 % with corresponding background probabilities of 2 $\\times$ 10$^{-4}$ and 3 $\\times$ 10$^{-2}$. This newly developed technique may enable water Cherenkov detectors to identify $\\bar \

Radiation portal monitors used for interdiction of illicit materials at borders include highly sensitive neutron detection systems. The main reason for having neutron detection capability is to detect fission neutrons from plutonium. The currently deployed radiation portal monitors (RPMs) from Ludlum and Science Applications International Corporation (SAIC) use neutrondetectors based upon 3He-filled gas proportional counters, which are the most common large neutrondetector. There is a declining supply of 3He in the world and, thus, methods to reduce the use of this gas in RPMs with minimal changes to the current system designs and sensitivity to cargo-borne neutrons are being investigated. Four technologies have been identified as being currently commercially available, potential alternative neutrondetectors to replace the use of 3He in RPMs. Reported here are the results of tests of the lithium-loaded glass fibers option. This testing measured the neutron detection efficiency and gamma ray rejection capabilities of a small system manufactured by Nucsafe (Oak Ridge, TN).

Radiation portal monitors used for interdiction of illicit materials at borders include highly sensitive neutron detection systems. The main reason for having neutron detection capability is to detect fission neutrons from plutonium. The currently deployed radiation portal monitors (RPMs) from Ludlum and Science Applications International Corporation (SAIC) use neutrondetectors based upon 3He-filled gas proportional counters, which are the most common large neutrondetector. There is a declining supply of 3He in the world, and thus, methods to reduce the use of this gas in RPMs with minimal changes to the current system designs and sensitivity to cargo-borne neutrons are being investigated. Four technologies have been identified as being currently commercially available, potential alternative neutrondetectors to replace the use of 3He in RPMs. Reported here are the results of tests of a newly designed boron-lined proportional counter option. This testing measured the neutron detection efficiency and gamma ray rejection capabilities of a system manufactured by Reuter Stokes.

PNNL-18938 Revision Radiation portal monitors used for interdiction of illicit materials at borders include highly sensitive neutron detection systems. The main reason for having neutron detection capability is to detect fission neutrons from plutonium. The currently deployed radiation portal monitors (RPMs) from Ludlum and Science Applications International Corporation (SAIC) use neutrondetectors based upon 3He-filled gas proportional counters, which are the most common large neutrondetector. There is a declining supply of 3He in the world, and thus, methods to reduce the use of this gas in RPMs with minimal changes to the current system designs and sensitivity to cargo-borne neutrons are being investigated. Four technologies have been identified as being currently commercially available, potential alternative neutrondetectors to replace the use of 3He in RPMs. Reported here are the results of tests of a newly designed boron-lined proportional counter option. This testing measured the neutron detection efficiency and gamma ray rejection capabilities of two successive prototypes of a system manufactured by GE Reuter Stokes.

1. The method of operating a water-cooled neutronic reactor having a graphite moderator which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40-60 volume percent of the mixture, in contact with the graphite moderator.

Two detectors for energy-resolved fast-neutron imaging in pulsed broad-energy neutron beams are presented. The first one is a neutron-counting detector based on a solid neutron converter coupled to a gaseous electron multiplier (GEM). The second is an integrating imaging technique, based on a scintillator for neutron conversion and an optical imaging system with fast framing capability.

Sample records for thermal neutron detector from the National Library of Energy Beta (NLEBeta)

Note: This page contains sample records for the topic "thermal neutron detector" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.

The TEMPEST II neutronthermalization code in Fortran for IBM 709 or 7090 calculates thermalneutron flux spectra based upon the Wigner-Wilkins equation, the Wilkins equation, or the Maxwellian distribution. When a neutron spectrum is obtained, TEMPEST II provides microscopic and macroscopic cross section averages over that spectrum. Equations used by the code and sample input and output data are given. (auth)

A Geant4-based Python/C++ simulation and coding framework, which has been developed and used in order to aid the R&D efforts for thermalneutrondetectors at neutron scattering facilities, is described. Built upon configurable geometry and generator modules, it integrates a general purpose object oriented output file format with meta-data, developed in order to facilitate a faster turn-around time when setting up and analysing simulations. Also discussed are the extensions to Geant4 which have been implemented in order to include the effects of low-energy phenomena such as Bragg diffraction in the polycrystalline support materials of the detector. Finally, an example application of the framework is briefly shown.

A neutrondetector has been constructed and tested at Pacific Northwest Laboratory (PNL) for the purpose of making fast, high sensitivity measurements of neutron emitters in portal applications. The system is based upon glass fiber optic scintillators loaded with lithium-6 and operated to detect thermalneutrons. Due to their compact size. physical flexibility, freedom from microphonic pickup, and complete lack of environmental and safety concerns, these fibers are very suitable for some applications. The electronics needed for these fibers is somewhat more complex than for helium-3 proportional counters, but the entire electronics package (including the controlling computer) has been shrunk into a space of 20 {times} 25 {times} 2 cm. The prototype sensor is about 180 {times} 60 {times} 7 cm, but a final design now under construction measures 200 {times} 28 {times} 2.54 cm. The new, smaller detectors will be capable of ganging to achieve any needed sensitivity and will each weigh about 16 kg. The principles of operation of the fiber will be discussed as will the operational mode of the detector.

A thermalneutron shield comprising concrete with a high percentage of the element Boron. The concrete is least 54% Boron by weight which maximizes the effectiveness of the shielding against thermalneutrons. The accompanying method discloses the manufacture of Boron loaded concrete which includes enriching the concrete mixture with varying grit sizes of Boron Carbide.

A borated polyethylene plane placed between a neutron source and a gamma spectrometer is used to form a dual neutron-gamma detection system. The polyethylene thermalizes the source neutrons so that they are captured by {sup 10}B to produce a flux of 478 keV gamma-rays that radiate from the plane. This results in a buildup of count rate in the detector over that from a disk of the same diameter as the detector crystal (same thickness as the panel). Radiation portal systems are a potential application of this technique.

A method of analysing Ge(Li) thermalneutron capture gamma spectra to obtain total gamma yields has been developed. Tie method determines both the yields from the well resolved gamma peaks in a spectrum as well as the gamma ...

Measuring alpha particles from a neutron induced break-up reaction with a mass spectrometer can be an excellent tool for detecting neutrons in a high neutron flux environment. Break-up reactions of {sup 6}Li and {sup 12}C can be used in the detection of slow and fast neutrons, respectively. A high neutron flux detection system that integrates the neutron energy sensitive material and helium mass spectrometer has been developed. The description of the detector configuration is given and it is soon to be tested at iThemba LABS, South Africa.

This project designed and built compound refractive lenses (CRLs) that are able to focus, collimate and image using thermalneutrons. Neutrons are difficult to manipulate compared to visible light or even x rays; however, CRLs can provide a powerful tool for focusing, collimating and imaging neutrons. Previous neutron CRLs were limited to long focal lengths, small fields of view and poor resolution due to the materials available and manufacturing techniques. By demonstrating a fabrication method that can produce accurate, small features, we have already dramatically improved the focal length of thermalneutron CRLs, and the manufacture of Fresnel lens CRLs that greatly increases the collection area, and thus efficiency, of neutron CRLs. Unlike a single lens, a compound lens is a row of N lenslets that combine to produce an N-fold increase in the refraction of neutrons. While CRLs can be made from a variety of materials, we have chosen to mold Teflon lenses. Teflon has excellent neutron refraction, yet can be molded into nearly arbitrary shapes. We designed, fabricated and tested Teflon CRLs for neutrons. We demonstrated imaging at wavelengths as short as 1.26 ? with large fields of view and achieved resolution finer than 250 ?m which is better than has been previously shown. We have also determined designs for Fresnel CRLs that will greatly improve performance.

Two legal-weight truck casks the GA-4 and GA-9, will carry four PWR and nine BWR spent fuel assemblies, respectively. Each cask has a solid neutron shielding material separating the steel body and the outer steel skin. In the thermal accident specified by NRC regulations in 10CFR Part 71, the cask is subjected to an 800[degree]C environment for 30 minutes. The neutron shield need not perform any shielding function during or after the thermal accident, but its behavior must not compromise the ability of the cask to contain the radioactive contents. In May-June 1989 the first series of full-scale thermal tests was performed on three shielding materials: Bisco Products NS-4-FR, and Reactor Experiments RX-201 and RX-207. The tests are described in Thermal Testing of Solid Neutron Shielding Materials, GA-AL 9897, R. H. Boonstra, General Atomics (1990), and demonstrated the acceptability of these materials in a thermal accident. Subsequent design changes to the cask rendered these materials unattractive in terms of weight or adequate service temperature margin. For the second test series, a material specification was developed for a polypropylene based neutron shield with a softening point of at least 280[degree]F. The neutron shield materials tested were boronated (0.8--4.5%) polymers (polypropylene, HDPE, NS-4). The Envirotech and Bisco materials are not polypropylene, but were tested as potential backup materials in the event that a satisfactory polypropylene could not be found.

The GERmanium Detector Array, GERDA, is designed for the search for ``neutrinoless double beta decay'' (0-nu-2-beta) with germanium detectors enriched in Ge76. An 18-fold segmented prototype detector for GERDA Phase II was exposed to an AmBe neutron source to improve the understanding of neutron induced backgrounds. Neutron interactions with the germanium isotopes themselves and in the surrounding materials were studied. Segment information is used to identify neutron induced peaks in the recorded energy spectra. The Geant4 based simulation package MaGe is used to simulate the experiment. Though many photon peaks from germanium isotopes excited by neutrons are correctly described by Geant4, some physics processes were identified as being incorrectly treated or even missing.

We present a detailed study of the spatial resolution of our time-resolved neutron imaging detector utilizing a new neutron position reconstruction method that improves both spatial resolution and event reconstruction efficiency. Our prototype detector system, employing a micro-pattern gaseous detector known as the micro-pixel chamber ({\\mu}PIC) coupled with a field-programmable-gate-array-based data acquisition system, combines 100{\\mu}m-level spatial and sub-{\\mu}s time resolutions with excellent gamma rejection and high data rates, making it well suited for applications in neutron radiography at high-intensity, pulsed neutron sources. From data taken at the Materials and Life Science Experimental Facility within the Japan Proton Accelerator Research Complex (J-PARC), the spatial resolution was found to be approximately Gaussian with a sigma of 103.48 +/- 0.77 {\\mu}m (after correcting for beam divergence). This is a significant improvement over that achievable with our previous reconstruction method (334 +/- 13 {\\mu}m), and compares well with conventional neutron imaging detectors and with other high-rate detectors currently under development. Further, a detector simulation indicates that a spatial resolution of less than 60 {\\mu}m may be possible with optimization of the gas characteristics and {\\mu}PIC structure. We also present an example of imaging combined with neutron resonance absorption spectroscopy.

Radiation portals normally incorporate a dedicated neutron counter and a gamma-ray detector with at least some spectroscopic capability. This paper describes the design and presents characterization data for a detection system called PVT-NG, which uses large polyvinyl toluene (PVT) detectors to monitor both types of radiation. The detector material is surrounded by polyvinyl chloride (PVC), which emits high-energy gamma rays following neutron capture reactions. Assessments based on high-energy gamma rays are well suited for the detection of neutron sources, particularly in border security applications, because few isotopes in the normal stream of commerce have significant gamma ray yields above 3 MeV. Therefore, an increased count rate for high-energy gamma rays is a strong indicator for the presence of a neutron source. The sensitivity of the PVT-NG sensor to bare {sup 252}Cf is 1.9 counts per second per nanogram (cps/ng) and the sensitivity for {sup 252}Cf surrounded by 2.5 cm of polyethylene is 2.3 cps/ng. The PVT-NG sensor is a proof-of-principal sensor that was not fully optimized. The neutrondetector sensitivity could be improved, for instance, by using additional moderator. The PVT-NG detectors and associated electronics are designed to provide improved resolution, gain stability, and performance at high-count rates relative to PVT detectors in typical radiation portals. As well as addressing the needs for neutron detection, these characteristics are also desirable for analysis of the gamma-ray spectra. Accurate isotope identification results were obtained despite the common impression that the absence of photopeaks makes data collected by PVT detectors unsuitable for spectroscopic analysis. The PVT detectors in the PVT-NG unit are used for both gamma-ray and neutron detection, so the sensitive volume exceeds the volume of the detection elements in portals that use dedicated components to detect each type of radiation.

The main objectives of this research project was to assemble, operate, test and characterize an innovatively designed scintillating fiber optic neutron radiation detector manufactured by Innovative American Technology with possible application to the Department of Homeland Security screening for potential radiological and nuclear threats at US borders (Kouzes 2004). One goal of this project was to make measurements of the neutron ship effect for several materials. The Virginia State University DOE FaST/NSF summer student-faculty team made measurements with the fiber optic radiation detector at PNNL above ground to characterize the ship effect from cosmic neutrons, and underground to characterize the muon contribution.

, where the energy generated is determined from measurements of heat balance. The lat- ter includes by standard methods of radiation transport, in particular with Monte Carlo methods. The fluid dynamic part are equivalent regarding their ability to account for the ef- fect of fluid dynamics on the detector time

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Radiation portal monitors used for interdiction of illicit materials at borders include highly sensitive neutron detection systems. The main reason for having neutron detection capability is to detect fission neutrons from plutonium. The currently deployed radiation portal monitors (RPMs) from Ludlum and Science Applications International Corporation (SAIC) use neutrondetectors based upon 3He-filled gas proportional counters, which are the most common large neutrondetector. There is a declining supply of 3He in the world, and thus, methods to reduce the use of this gas in RPMs with minimal changes to the current system designs and sensitivity to cargo-borne neutrons are being investigated. Four technologies have been identified as being currently commercially available, potential alternative neutrondetectors to replace the use of 3He in RPMs. Reported here are the results of tests of a boron-lined proportional counter design variation. In the testing described here, the neutron detection efficiency and gamma ray rejection capabilities of a system manufactured by Proportional Technologies, Inc, was tested.

Two acrylic cube phantoms have been constructed for BNCT applications that allow the depth distribution of neutrons to be measured with miniature {sup 10}BF{sub 3} detectors in 0.5-cm steps beginning at 1-cm depth. Sizes and weights of the cubes are 14 cm, 3.230 kg, and 11 cm, 1.567 kg. Tests were made with the epithermal neutron beam from the patient treatment port of the Brookhaven Medical Research Reactor. Thermalneutron depth profiles were measured with a bare {sup 10}BF{sub 3} detector at a reactor power of 50 W, and Cd-covered detector profiles were measured at a reactor power of 1 kW. The resulting plots of counting rate versus depth illustrate the dependence of neutron moderation on the size of the phantom. But more importantly the data can serve as benchmarks for testing the thermal and epithermal neutron profiles obtained with accelerator-based BNCT facilities. Such tests could be made with these phantoms at power levels about five orders of magnitude lower than that required for the treatment of patients with brain tumors. {copyright} {ital 1998 American Association of Physicists in Medicine.}

Effective detection of special nuclear materials (SNM) is essential for reducing the threat associated with stolen or improvised nuclear devices. Passive radiation detection technologies are primarily based on gamma-ray detection and subsequent isotope identification or neutron detection (specific to neutron sources and SNM). One major effort supported by the Department of Homeland Security in the area of advanced passive detection is handheld or portable neutrondetectors for search and localization tasks in emergency response and interdiction settings. A successful SNM search detector will not only be able to confirm the presence of fissionable materials but also establish the location of the source in as short of time as possible while trying to minimize false alarms due to varying background or naturally occurring radioactive materials (NORM). For instruments based on neutrondetectors, this translates to detecting neutrons from spontaneous fission or alpha-n reactions and being able to determine the direction of the source (or localizing the source through subsequent measurements). Polyethylene-coated gallium arsenide detectors were studied because the detection scheme is based on measuring the signal in the gallium arsenide wafers from the electrical charge of the recoil protons produced from the scattering of neutrons from the hydrogen nucleus. The inherent reaction has a directional dependence because the neutron and hydrogen nucleus have equivalent masses. The assessment and measurement of polyethylene-coated gallium arsenide detector properties and characteristics was the first phase of a project being performed for the Department of Homeland Security and the results of these tests are reported in this report. The ultimate goal of the project was to develop a man-portable neutron detection system that has the ability to determine the direction of the source from the detector. The efficiency of GaAs detectors for different sizes of polyethylene layers and different angles between the detector and the neutron source were determined. Preliminary measurements with a neutron generator based on a deuterium-tritium reaction ({approx}14 MeV neutrons) were performed and the results are discussed. This report presents the results of these measurements in terms of efficiency and angular efficiency and compares them to Monte Carlo calculations to validate the calculation scheme in view of further applications. Based on the results of this study, the polyethylene-coated gallium arsenide detectors provide adequate angular resolution based on proton recoil detection from the neutron scattering reaction from hydrogen. However, the intrinsic efficiency for an individual detector is extremely low. Because of this low efficiency, large surface area detectors ( or a large total surface area from many small detectors) would be required to generate adequate statistics to perform directional detection in near-real time. Large surface areas could be created by stacking the detector wafers with only a negligible attenuation of source neutrons. However, the cost of creating such a large array of GaAs is cost-prohibitive at this time.

A nuclear reactor for the irradiation of materials designed to provide a localized area of high thermalneutron flux density in which the materials to be irradiated are inserted is described. The active portion of the reactor is comprised of a cubicle graphite moderator of about 25 feet in length along each axis which has a plurality of cylindrical channels for accommodatirg elongated tubular-shaped fuel elements. The fuel elements have radial fins for spacing the fuel elements from the channel walls, thereby providing spaces through which a coolant may be passed, and also to serve as a heatconductirg means. Ducts for accommnodating the sample material to be irradiated extend through the moderator material perpendicular to and between parallel rows of fuel channels. The improvement is in the provision of additional fuel element channels spaced midway between 2 rows of the regular fuel channels in the localized area surrounding the duct where the high thermalneutron flux density is desired. The fuel elements normally disposed in the channels directly adjacent the duct are placed in the additional channels, and the channels directly adjacent the duct are plugged with moderator material. This design provides localized areas of high thermalneutron flux density without the necessity of providing additional fuel material.

Radiation portal monitors used for interdiction of illicit materials at borders include highly sensitive neutron detection systems. There is a declining supply of 3He in the world, and thus, methods to reduce the use of this gas in RPMs with minimal changes to the current system designs and sensitivity to cargo-borne neutrons are being investigated. Four technologies have been identified as being currently commercially available, potential alternative neutrondetectors to replace the use of 3He in RPMs. In addition, a few other companies have detector technologies that might be competitive in the near term as an alternative technology. Reported here are the results of tests of 6Li/ZnS(Ag)-coated scintillator paddles. This testing measured the required performance for neutron detection efficiency and gamma ray rejection capabilities of a system manufactured by Symetrica.

Fusion reactions in an inertial-confinement fusion (ICF) target filled with deuterium or a deuterium/tritium fuel release nearly monoenergetic neutrons. Because most the neutrons leave the compressed target without collision, they preserve reaction-rate information as they travel radially outward from their point of origin. Three fast, neutrondetector techniques, each capable of measuring the fusion reaction-rate of ICF targets, have been demonstrated. The most advanced detector is based on the fast rise-time of a commercial plastic scintillator material (BC-422) which acts as a neutron-to-light converter. Signals, which are recorded with a fast optical streak camera, have a resolution of 25 ps. Good signals can be recorded for targets producing only 5 x 10{sup 7} DT neutrons. Two other detectors use knock-on collisions between neutrons and protons in a thin polyethylene (CH{sub 2}) converter. In one, the converter is placed in front of the photocathode of an x-ray streak camera. Recoil protons pass through the photocathode and knock out electrons which are accelerated and deflected to produce a signal. Resolutions < 25 ps are possible. In the other, the converter is placed in front of a microchannel plate (MCP) with a gated microstrip. Recoil protons eject electrons from the gold layer forming the microstrip. If a gate pulse is present, the signal is amplified. Present gate times are about 80 ps.

Current Joint Test Assembly (JTA) neutron monitors rely on knock-on proton type detectors that are susceptible to X-rays and low energy gamma rays. We investigated two novel plastic scintillating fiber directional neutrondetector prototypes. One prototype used a fiber selected such that the fiber width was less than 2.1mm which is the range of a proton in plastic. The difference in the distribution of recoil proton energy deposited in the fiber was used to determine the incident neutron direction. The second prototype measured both the recoil proton energy and direction. The neutron direction was determined from the kinematics of single neutron-proton scatters. This report describes the development and performance of these detectors.

A gamma-free neutron-sensitive scintillator is needed to enhance radiaition sensing and detection for nonproliferation applications. Such a scintillator would allow very large detectors to be placed at the perimeter of spent-fuel storage facilities at commercial nuclear power plants, so that any movement of spontaneously emitted neutrons from spent nuclear fuel or weapons grade plutonium would be noted in real-time. This task is to demonstrate that the technology for manufacturing large panels of fluor-doped plastic containing lithium-6 phosphate nanoparticles can be achieved. In order to detect neutrons, the nanoparticles must be sufficiently small so that the plastic remains transparent. In this way, the triton and alpha particles generated by the capture of the neutron will result in a photon burst that can be coupled to a wavelength shifting fiber (WLS) producing an optical signal of about ten nanoseconds duration signaling the presence of a neutron emitting source.

Spontaneous and induced fission in Special Nuclear Material (SNM) such as 235U and 239Pu results in the emission of neutrons and high energy gamma-rays. The multiplicities of and time correlations between these particles are both powerful indicators of the presence of fissile material. Detectors sensitive to these signatures are consequently useful for nuclear material monitoring, search, and characterization. In this article, we demonstrate sensitivity to both high energy gamma-rays and neutrons with a water Cerenkov based detector. Electrons in the detector medium, scattered by gamma-ray interactions, are detected by their Cerenkov light emission. Sensitivity to neutrons is enhanced by the addition of a gadolinium compound to the water in low concentrations. Cerenkov light is similarly produced by an 8 MeV gamma-ray cascade following neutron capture on the gadolinium. The large solid angle coverage and high intrinsic efficiency of this detection approach can provide robust and low cost neutron and gamma-ray detection with a single device.

A test campaign was undertaken during April 16-19 in LaHonda, California to match the operational performance of the Idaho National Engineering and Environmental Laboratory (INEEL)Varitron accelerator to that of an ARACOR Eagle accelerator. This Eagle-matched condition, with the INEEL Varitron, will be used during a concept demonstration test at Los Alamos National Laboratory (LANL). This operational characterization involved the use of similar electron beam energies, similar production of photoneutrons from selected non-nuclear materials, and similar production of photofissionbased, delayed neutrons from an INEEL-provided, depleted uranium sample. Then using the matched operation, the Varitron was used to define detector performances for several INEEL and LANL detectors using the depleted uranium target and Eagle-like, bremsstrahlung collimation. This summary report provides neutron measurements using the INEEL detectors. All delayed neutron data are acquired in the time interval ranging from 4.95 to 19.9 ms after each accelerator pulse. All prompt neutron data are acquired during 0.156 to 4.91 ms after each accelerator pulse. Prompt and delayed neutron counting acquisition intervals can still be optimized.

Recent program requirements of the US Department of Energy/NNSA have led to a need for a criticality accident alarm system to be installed at a newly activated facility. The Criticality Safety Group of the Lawrence Livermore National Laboratory (LLNL) was able to recover and store for possible future use approximately 200 neutron criticality detectors and 20 master alarm panels from the former Rocky Flats Plant in Golden, Colorado when the plant was closed. The Criticality Safety Group participated in a facility analysis and evaluation, the engineering design and review process, as well as the refurbishment, testing, and recalibration of the Rocky Flats criticality alarm system equipment to be used in the new facility. In order to demonstrate the functionality and survivability of the neutrondetectors to the effects of an actual criticality accident, neutrondetector testing was performed at the French CEA Valduc SILENE reactor from October 7 to October 19, 2010. The neutrondetectors were exposed to three criticality events or pulses generated by the SILENE reactor. The first excursion was performed with a bare or unshielded reactor, and the second excursion was made with a lead shielded/reflected reactor, and the third excursion with a polyethylene reflected core. These tests of the Rocky Flats neutrondetectors were performed as a part of the 2010 Criticality Accident Alarm System Benchmark Measurements at the SILENE Reactor. The principal investigators for this series of experiments were Thomas M. Miller and John C. Wagner of the Oak Ridge National Laboratory, with Nicolas Authier and Nathalie Baclet of CEA Valduc. Several other organizations were also represented, including the Y-12 National Security Complex, Lawrence Livermore National Laboratory, Los Alamos National Laboratory, CEA Saclay, and Babcock International Group.

Neutron scattering is a powerful technique that is critically important for materials science and structural biology applications. The knowledge gained from past developments has resulted in far-reaching advances in engineering, pharmaceutical and biotechnology industries, to name a few. New facilities for neutron generation at much higher flux, such as the SNS at Oak Ridge, TN, will greatly enhance the capabilities of neutron scattering, with benefits that extend to many fields and include, for example, development of improved drug therapies and materials that are stronger, longer-lasting, and more impact-resistant. In order to fully realize this enhanced potential, however, higher neutron rates must be met with improved detection capabilities, particularly higher count rate capability in large size detectors, while maintaining practicality. We have developed a neutrondetector with the technical and economic advantages to accomplish this goal. This new detector has a large sensitive area, offers 3D spatial resolution, high sensitivity and high count rate capability, and it is economical and practical to produce. The proposed detector technology is based on B-10 thin film conversion of neutrons in long straw-like gas detectors. A stack of many such detectors, each 1 meter in length, and 4 mm in diameter, has a stopping power that exceeds that of He-3 gas, contained at practical pressures within an area detector. With simple electronic readout methods, straw detector arrays can provide spatial resolution of 4 mm FWHM or better, and since an array detector of such form consists of several thousand individual elements per square meter, count rates in a 1 m^2 detector can reach 2?10^7 cps. Moreover, each individual event can be timetagged with a time resolution of less than 0.1 ?sec, allowing accurate identification of neutron energy by time of flight. Considering basic elemental cost, this novel neutron imaging detector can be commercially produced economically, probably at a small fraction of the cost of He-3 detectors. In addition to neutron scattering science, the fully developed base technology can be used as a rugged, low-cost neutrondetector in area monitoring and surveying. Radiation monitors are used in a number of other settings for occupational and environmental radiation safety. Such a detector can also be used in environmental monitoring and remote nuclear power plant monitoring. For example, the Department of Energy could use it to characterize nuclear waste dumps, coordinate clean-up efforts, and assess the radioactive contaminants in the air and water. Radiation monitors can be used to monitor the age and component breakdown of nuclear warheads and to distinguish between weapons and reactor grade plutonium. The UN's International Atomic Energy Agency (IAEA) uses radiation monitors for treaty verification, remote monitoring, and enforcing the non-proliferation of nuclear weapons. As part of treaty verification, monitors can be used to certify the contents of containers during inspections. They could be used for portal monitoring to secure border checkpoints, sea ports, air cargo centers, public parks, sporting venues, and key government buildings. Currently, only 2% of all sea cargo shipped is inspected for radiation sources. In addition, merely the presence of radiation is detected and nothing is known about the radioactive source until further testing. The utilization of radiation monitors with neutron sensitivity and capability of operation in hostile port environments would increase the capacity and effectiveness of the radioactive scanning processes.

With an increase in the capabilities and sophistication of terrorist networks worldwide comes a corresponding increase in the probability of a radiological or nuclear device being detonated within the borders of the United States. One method to decrease the risk associated with this threat is to interdict the material during transport into the US. Current RPMS have limitations in their ability to detect shielded nuclear materials. It was proposed that directionally sensitive neutrondetectors might be able to overcome many of these limitations. This thesis presents a method to create a directionally sensitive neutrondetector using a unique characteristic of 10B. This characteristic is the Doppler broadening of the de-excitation gamma-ray from the 10B(n, alpha) reaction. Using conservation principles and the method of cone superposition, the mathematics for determining the incoming neutron direction vector from counts in a boron loaded cloud chamber and boron loaded semiconductor were derived. An external routine for MCNPX was developed to calculate the Doppler broaden de-excitation gamma-rays. The calculated spectrum of Doppler broadened de-excitation gamma-rays was then compared to measured and analytical spectrums and matched with a high degree of accuracy. MCNPX simulations were performed for both a prototype 10B loaded cloud chamber and prototype 10B loaded semiconductor detector. These simulations assessed the detectors' abilities to determine incoming neutron direction vectors using simulated particle reactant data. A sensitivity analysis was also performed by modifying the energy and direction vector of the simulated output data for 7Li* particles. Deviation coefficients showed a respective angular uncertainty of 1.86 degrees and 6.07 degrees for the boron loaded cloud chamber and a boron loaded semiconductor detectors. These capabilities were used to propose a possible RPM design that could be implemented.

Three types of neutrondetectors (plastic scintillation detectors, indium activation detectors, and CR-39 track detectors) were calibrated for the measurement of 2.45 MeV DD fusion neutron yields from the deuterium cluster fusion experiment on the Texas Petawatt Laser. A Cf-252 neutron source and 2.45 MeV fusion neutrons generated from laser-cluster interaction were used as neutron sources. The scintillation detectors were calibrated such that they can detect up to 10{sup 8} DD fusion neutrons per shot in current mode under high electromagnetic pulse environments. Indium activation detectors successfully measured neutron yields as low as 10{sup 4} per shot and up to 10{sup 11} neutrons. The use of a Cf-252 neutron source allowed cross calibration of CR-39 and indium activation detectors at high neutron yields ({approx}10{sup 11}). The CR-39 detectors provided consistent measurements of the total neutron yield of Cf-252 when a modified detection efficiency of 4.6 Multiplication-Sign 10{sup -4} was used. The combined use of all three detectors allowed for a detection range of 10{sup 4} to 10{sup 11} neutrons per shot.

This work presents the results obtained in the inspection of several mechanical components through neutron and gamma-ray transmission radiography. The 4.46 Multiplication-Sign 10{sup 5} n.cm{sup -2}.s{sup -1} thermalneutron flux available at the main port of the Argonauta research reactor in Instituto de Engenharia Nuclear has been used as source for the neutron radiographic imaging. The 412 keV {gamma}-ray emitted by {sup 198}Au, also produced in that reactor, has been used as interrogation agent for the gamma radiography. Imaging Plates - IP specifically designed to operate with thermalneutrons or with X-rays have been employed as detectors and storage devices for each of these radiations.

Neutron And Gamma Detector Using An Ionization Chamber Neutron And Gamma Detector Using An Ionization Chamber Neutron And Gamma Detector Using An Ionization Chamber With An Integrated Body And Moderator A detector for detecting neutrons and gamma radiation includes a cathode that defines an interior surface and an interior volume. Available for thumbnail of Feynman Center (505) 665-9090 Email Neutron And Gamma Detector Using An Ionization Chamber With An Integrated Body And Moderator A detector for detecting neutrons and gamma radiation includes a cathode that defines an interior surface and an interior volume. A conductive neutron-capturing layer is disposed on the interior surface of the cathode and a plastic housing surrounds the cathode. A plastic lid is attached to the housing and encloses the interior volume of the cathode forming an

Non-streaming high-efficiency perforated semiconductor neutrondetectors, method of making same and measuring wands and detector modules utilizing same are disclosed. The detectors have improved mechanical structure, flattened angular detector responses, and reduced leakage current. A plurality of such detectors can be assembled into imaging arrays, and can be used for neutron radiography, remote neutron sensing, cold neutron imaging, SNM monitoring, and various other applications.

A test program of the performance of 3He neutron proportional detectors with varying gas pressures, and their response to lligh level gamma-ray exposure in a mixed neutrodgamma environment, ha$ been performed Our intent was to identie the optimal gas pressure to reduce the gamma-ray sensitivity of these detectors. These detectors were manufxtured using materials to minimize their gamma response. Earlier work focused on 3He fill pressures of four atmospheres and above, whereas the present work focuses on a wider range of pressures. Tests have shown that reducing the .filling pressure will M e r increase the gamma-ray dose range in which the detectors can be operated.

Sample records for thermal neutron detector from the National Library of Energy Beta (NLEBeta)

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In order to clarify the effects of fusion-produced neutron irradiation on silicon semiconductor x-ray detectors, the x-ray energy responses of both n- and p-type silicon tomography detectors used in the Joint European Torus (JET) tokamak (n-type) and the GAMMA 10 tandem mirror (p-type) are studied using synchrotron radiation at the Photon Factory of the National Laboratory for High Energy Accelerator Research Organization (KEK). The fusion neutronics source (FNS) of Japan Atomic Energy Research Institute (JAERI) is employed as well-calibrated D-T neutron source with fluences from 10{sup 13} to 10{sup 15} neutrons/cm{sup 2} onto these semiconductor detectors. Different fluence dependence is found between these two types of detectors; that is, (i) for the n-type detector, the recovery of the degraded response is found after the neutron exposure beyond around 10{sup 13} neutrons/cm{sup 2} onto the detector. A further finding is followed as a 're-degradation' by a neutron irradiation level over about 10{sup 14} neutrons/cm{sup 2}. On the other hand, (ii) the energy response of the p-type detector shows only a gradual decrease with increasing neutron fluences. These properties are interpreted by our proposed theory on semiconductor x-ray responses in terms of the effects of neutrons on the effective doping concentration and the diffusion length of a semiconductor detector.

Prompt fission neutron spectrum measurements at the University of Massachusetts Lowell 5.5 MV Van de Graaff accelerator laboratory require that the neutrondetector efficiency be well known over a neutron energy range of 100 keV to 20 MeV. The efficiency of the detector, has been determined for energies greater than 5.0 MeV using the Weapons Neutron Research (WNR) white neutron source at the Los Alamos Meson Physics Facility (LAMPF) in a pulsed beam, time-of-flight (TOF) experiment. Carbon matched polyethylene and graphite scatterers were used to obtain a hydrogen spectrum. The detector efficiency was determined using the well known H(n,n) scattering cross section. Results are compared to the detector efficiency calculation program SCINFUL available from the Radiation Shielding Information Center at Oak Ridge National Laboratory.

Collaboration between the Pacific Northwest National Laboratory (PNNL) and the Los Alamos National Laboratory (LANL) is underway to evaluate neutron detection technologies that might replace the high-pressure helium (3He) tubes currently used in neutron multiplicity counter for safeguards applications. The current stockpile of 3He is diminishing and alternatives are needed for a variety of neutron detection applications including multiplicity counters. The first phase of this investigation uses a series of Monte Carlo calculations to simulate the performance of an existing neutron multiplicity counter configuration by replacing the 3He tubes in a model for that counter with candidate alternative technologies. These alternative technologies are initially placed in approximately the same configuration as the 3He tubes to establish a reference level of performance against the 3He-based system. After these reference-level results are established, the configurations of the alternative models will be further modified for performance optimization. The 3He model for these simulations is the one used by LANL to develop and benchmark the Epithermal Neutron Multiplicity Counter (ENMC) detector, as documented by H.O. Menlove, et al. in the 2004 LANL report LA-14088. The alternative technologies being evaluated are the boron-tri-fluoride-filled proportional tubes, boron-lined tubes, and lithium coated materials previously tested as possible replacements in portal monitor screening applications, as documented by R.T. Kouzes, et al. in the 2010 PNNL report PNNL-72544 and NIM A 623 (2010) 10351045. The models and methods used for these comparative calculations will be described and preliminary results shown

The authors recently designed and built a compact neutrondetector for a geology experiment. The detector had to fit inside a 1.5-in.-diam borehole in a large block of concrete. They attached a gas-filled, 1-in.-diam {sup 3}He tube to a 1-in.-diam electronics preamplifier package of their design. The electronics package consists of a cylindrically shaped, high-voltage section and a single-channel analyzer with a buffered output. The low-voltage components are mounted on a printed-circuit board. The circuit board and the high-voltage section are attached to a semicylindrical base. The outputs consist of a light-emitting diode for visual observations and a fixed-width, TTL-compatible pulse for a counter. This internal assembly is equipped with coaxial connectors and slips into a thin-walled tube that serves as the preamplifier housing. Power for a detector is supplied by an external, high-voltage supply and a 5-Vdc supply.

The first several campaigns of laser fusion experiments at the National Ignition Facility (NIF) included a family of high-sensitivity scintillator/photodetector neutron-time-of-flight (nTOF) detectors for measuring deuterium-deuterium (DD) and DT neutron yields. The detectors provided consistent neutron yield (Y{sub n}) measurements from below 10{sup 9} (DD) to nearly 10{sup 15} (DT). The detectors initially demonstrated detector-to-detector Y{sub n} precisions better than 5%, but lacked in situ absolute calibrations. Recent experiments at NIF now have provided in situ DT yield calibration data that establish the absolute sensitivity of the 4.5 m differential tissue harmonic imaging (DTHI) detector with an accuracy of {+-}10% and precision of {+-}1%. The 4.5 m nTOF calibration measurements also have helped to establish improved detector impulse response functions and data analysis methods, which have contributed to improving the accuracy of the Y{sub n} measurements. These advances have also helped to extend the usefulness of nTOF measurements of ion temperature and downscattered neutron ratio (neutron yield 10-12 MeV divided by yield 13-15 MeV) with other nTOF detectors.

An initial study is performed to determine how temperature considerations affect LIFE neutronic simulations. Among other figures of merit, the isotopic mass accumulation, thermal power, tritium breeding, and criticality are analyzed. Possible fidelities of thermal modeling and degrees of coupling are explored. Lessons learned from switching and modifying nuclear datasets is communicated.

A new neutron time-of-flight (nTOF) detector with a bibenzyl crystal as a scintillator has been designed and manufactured for the National Ignition Facility (NIF). This detector will replace a nTOF20-Spec detector with an oxygenated xylene scintillator currently operational on the NIF to improve the areal-density measurements. In addition to areal density, the bibenzyl detector will measure the D-D and D-T neutron yield and the ion temperature of indirect- and direct-drive-implosion experiments. The design of the bibenzyl detector and results of tests on the OMEGA Laser System are presented.

The prompt gamma-ray data of thermal- neutron captures fornuclear mass number A=26-35 had been evaluated and published in "ATOMICDATA AND NUCLEAR DATA TABLES, 26, 511 (1981)". Since that time the manyexperimental data of the thermal-neutron captures have been measured andpublished. The update of the evaluated prompt gamma-ray data is verynecessary for use in PGAA of high-resolution analytical prompt gamma-rayspectroscopy. Besides, the evaluation is also very needed in theEvaluated Nuclear Structure Data File, ENSDF, because there are no promptgamma-ray data in ENSDF. The levels, prompt gamma-rays and decay schemesof thermal-neutron captures fornuclides (26Mg, 27Al, 28Si, 29Si, 30Si,31P, 32S, 33S, 34S, and 35Cl) with nuclear mass number A=26-35 have beenevaluated on the basis of all experimental data. The normalizationfactors, from which absolute prompt gamma-ray intensity can be obtained,and necessary comments are given in the text. The ENSDF format has beenadopted in this evaluation. The physical check (intensity balance andenergy balance) of evaluated thermal-neutron capture data has been done.The evaluated data have been put into Evaluated Nuclear Structure DataFile, ENSDF. This evaluation may be considered as an update of the promptgamma-ray from thermal-neutron capture data tables as published in"ATOMIC DATA AND NUCLEAR DATA TABLES, 26, 511 (1981)".

A neutrondetector based on EJ301 liquid scintillator has been employed at EAST to measure the neutron energy spectrum for D-D fusion plasma. The detector was carefully characterized in different quasi-monoenergetic neutron fields generated by a 4.5 MV Van de Graaff accelerator. In recent experimental campaigns, due to the low neutron yield at EAST, a new shielding device was designed and located as close as possible to the tokamak to enhance the count rate of the spectrometer. The fluence of neutrons and gamma-rays was measured with the liquid neutron spectrometer and was consistent with 3He proportional counter and NaI (Tl) gamma-ray spectrometer measurements. Plasma ion temperature values were deduced from the neutron spectrum in discharges with lower hybrid wave injection and ion cyclotron resonance heating. Scattered neutron spectra were simulated by the Monte Carlo transport Code, and they were well verified by the pulse height measurements at low energies.

of Physics and Power Engineering (Obninsk) have proposed an accelerator based neutron source for neutron for neutron capture therapy is under construction now at the Budker Institute of Nuclear Physics, NovosibirskAccelerator based neutron source for neutron capture therapy B. Bayanov, Yu. Belchenko, V. Belov, V

In this thesis, we designed and tested a calibration and deployment system for the MiniCLEAN dark matter detector. The deployment system uses a computer controlled winch to lower a canister containing a neutron source into ...

A solid state nuclear track detector, CR-39, was exposed to DT neutrons. After etching, the resultant tracks were analyzed using both an optical microscope and a scanning electron microscope (SEM). In this communication, both methods of analyzing DT neutron tracks are discussed.

The Versatile Array of NeutronDetectors at Low Energy (VANDLE) is a new array of plastic scintillator bars under development for measurements at the Holifield Radioactive Ion Beam Facility at Oak Ridge National Laboratory. The array is highly modular allowing the configuration of the individual elements to be optimized for particular experimental requirements, such as (d,n) and beta-delayed neutron-decay measurements with neutron-rich rare isotope beams.

The next generation of radioactive ion beam facilities, which will give experimental access to many exotic nuclei, are presently being developed. These facilities will make it possible to study very short lived exotic nuclei with extreme values of isospin far from the line of beta stability. Such nuclei will be produced with very low cross sections and to study them, new detector arrays are being developed. At the SPIRAL facility in GANIL a neutrondetector array, the Neutron Wall, is located. In this work the Neutron Wall has been characterized regarding neutron detection efficiency and discrimination between neutrons and gamma rays. The possibility to increase the efficiency by increasing the high voltage of the photomultiplier tubes has also been studied. For SPIRAL2 a neutrondetector array, NEDA, is being developed. NEDA will operate in a high gamma-ray background environment which puts a high demand on the quality of discrimination between neutrons and gamma rays. To increase the quality of the discrimination methods pulse-shape discrimination techniques utilizing digital electronics have been developed and evaluated regarding bit resolution and sampling frequency of the ADC. The conclusion is that an ADC with a bit resolution of 12 bits and a sampling frequency of 100 MS/s is adequate for pulse-shape discrimination of neutrons and gamma rays for a neutron energy range of 0.3-12 MeV.

Novel ultra-compact, electrically switchable, time-structured/pulsed, ~1-14 MeV-level neutron and photon generators have application embedded into large detector systems, especially calorimeters, for energy and operational calibration. The small sizes are applicable to permanent in-situ deployment, or able to be conveniently inserted into large high energy physics detector systems. For bench- testing of prototypes, or for detector module production testing, these compact n and gamma generators offer advantages.

Neutron depth profiling (NDP) is a mature, nondestructive technique used to characterize the concentration of certain light isotopes in a material as a function of depth by measuring the residual energy of charged particles in neutron induced reactions. Historically, NDP has been performed using a single detector, resulting in low intrinsic detection efficiency, and limiting the technique largely to high flux research reactors. In this work, we describe a new NDP instrument design with higher detection efficiency by way of spectrum summing across multiple detectors. Such a design is capable of acquiring a statistically significant charged particle spectrum at facilities limited in neutron flux and operation time.

The large difference in neutron scattering cross-section at low neutron energies between ortho- and para-hydrogen was recognized early on. In view of this difference (more than an order of magnitude), one might legitimately ask whether the ortho/para ratio has a significant effect on the neutronthermalization properties of a cold hydrogen moderator. Several experiments performed in the 60`s and early 70`s with a variety of source and (liquid hydrogen) moderator configurations attempted to investigate this. The results tend to show that the ortho/para ratio does indeed have an effect on the energy spectrum of the neutron beam produced. Unfortunately, the results are not always consistent with each other and much unknown territory remains to be explored. The problem has been approached from a computational standpoint, but these isolated efforts are far from having examined the ortho/para-hydrogen problem in neutron moderation in all its complexity. Because of space limitations, the authors cannot cover, even briefly, all the aspects of the ortho/para question here. This paper will summarize experiments meant to investigate the effect of the ortho/para ratio on the neutron energy spectrum produced by liquid hydrogen moderators.

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Various neutron diagnostic tools are used in fusion reactors to evaluate different aspects of plasma performance, such as fusion power, power density, ion temperature, fast ion energy, and their spatial distributions. The stilbene scintillator has been proposed for use as a neutron diagnostic system to measure the characteristics of neutrons from the Korea Superconducting Tokamak Advanced Research (KSTAR) fusion reactor. Specially designed electronics are necessary to measure fast neutron spectra with high radiation from a gamma-ray background. The signals from neutrons and gamma-rays are discriminated by the digital charge pulse shape discrimination (PSD) method, which uses total to partial charge ratio analysis. The signals are digitized by a flash analog-to-digital convertor (FADC). To evaluate the performance of the fabricated stilbene neutron diagnostic system, the efficiency of 10 mm soft-iron magnetic shielding and the detection efficiency of fast neutrons were tested experimentally using a 252Cf neutron source. In the results, the designed and fabricated stilbene neutron diagnostic system performed well in discriminating neutrons from gamma-rays under the high magnetic field conditions during KSTAR operation. Fast neutrons of 2.45 MeV were effectively measured and evaluated during the 2011 KSTAR campaign.

Various neutron diagnostic tools are used in fusion reactors to evaluate different aspects of plasma performance, such as fusion power, power density, ion temperature, fast ion energy, and their spatial distributions. The stilbene scintillator has been proposed for use as a neutron diagnostic system to measure the characteristics of neutrons from the Korea Superconducting Tokamak Advanced Research (KSTAR) fusion reactor. Specially designed electronics are necessary to measure fast neutron spectra with high radiation from a gamma-ray background. The signals from neutrons and gamma-rays are discriminated by the digital charge pulse shape discrimination (PSD) method, which uses total to partial charge ratio analysis. The signals are digitized by a flash analog-to-digital convertor (FADC). To evaluate the performance of the fabricated stilbene neutron diagnostic system, the efficiency of 10 mm soft-iron magnetic shielding and the detection efficiency of fast neutrons were tested experimentally using a 252Cf neutr...

This paper demonstrates, through MCNPX simulations, that a compact hexagonal array of detectors can be utilized to do both gamma isotopic identification (ID) along with neutron identification while simultaneously finding the direction of the source relative to the detector array. The detector array itself is composed of seven borated polyvinyl toluene (PVT) hexagonal light pipes approximately 4 inches long and with a 1.25 inch face-to-face thickness assembled in a tight configuration. The gamma ID capability is realized through judicious windowing algorithms as is the neutron spectral unfolding. By having multiple detectors in different relative positions, directional determination of the source can be realized. By further adding multiplicity counters to the neutron counts, fission events can be measured.

Experimental and theoretical work on the interaction mechanisms by which neutrons exchange energy with H atoms involves treating neutronthermalization as neutron interactions with energy levels in the atoms. Cold moderators are presently being studied in order to optimize the source of cold neutrons. Cold neutrons are provided from an accelerator arrangement that directs electrons against a Fansteel target producing fast neutrons. Thermalneutrons, produced by moderation of fast neutrons, are passed through a chopper. Several moderators are evaluated, and neutron emission time measurements by crystal diffraction and beam chopper techniques point out emission time dependence on thickness, moderator, and temperature. The neutron beam chopper used presently is described, and results of neutron scattering by liquid para- and orthohydrogen are displayed and compared with theoretical predictions made with a perfect hydrogen gas model. Inelastic scattering of neutrons by liquid H is discussed, and theoretical and experimental results of inelastic scattering by polyethylene are also included. (D.C.W.)

The event-by-event fission model FREYA is extended to spontaneous fission of actinides and a variety of neutron observables are studied for spontaneous fission and fission induced by thermalneutrons with a view towards possible applications for SNM detection. We have shown that event-by-event models of fission, such as FREYA, provide a powerful tool for studying fission neutron correlations. Our results demonstrate that these correlations are significant and exhibit a dependence on the fissioning nucleus. Since our method is phenomenological in nature, good input data are especially important. Some of the measurements employed in FREYA are rather old and statistics limited. It would be useful to repeat some of these studies with modern detector techniques. In addition, most experiments made to date have not made simultaneous measurements of the fission products and the prompt observables, such as neutron and photons. Such data, while obviously more challenging to obtain, would be valuable for achieving a more complete understanding of the fission process.

The next generation of radioactive ion beam facilities, which will give experimental access to many exotic nuclei, are presently being developed. These facilities will make it possible to study very short lived exotic nuclei with extreme values of isospin far from the line of beta stability. Such nuclei will be produced with very low cross sections and to study them, new detector arrays are being developed. At the SPIRAL facility in GANIL a neutrondetector array, the Neutron Wall, is located. In this work the Neutron Wall has been characterized regarding neutron detection efficiency and discrimination between neutrons and gamma rays. The possibility to increase the efficiency by increasing the high voltage of the photomultiplier tubes has also been studied. For SPIRAL2 a neutrondetector array, NEDA, is being developed. NEDA will operate in a high gamma-ray background environment which puts a high demand on the quality of discrimination between neutrons and gamma rays. To increase the quality of the discrimi...

The quiescent thermal emission from neutron stars in low mass X-ray binaries after active periods of intense activity in x-rays (outbursts) has been monitored. The theoretical modeling of the thermal relaxation of the neutron star crust may be used to establish constraints on the crust and envelope composition and transport properties, depending on the astrophysical scenarios assumed. We perform numerical simulations of the neutron star crust thermal evolution and compare them with inferred surface temperatures for five sources: MXB 1659-29, KS 1731-260, EXO 0748-676, XTE J1701-462 and IGR J17480-2446. We also present stationary envelope models to be used as a boundary condition for the crustal cooling models. We obtain a relation between the mass accretion rate and the temperature reached at the crust-envelope interface at the end of the active phase that accounts for early observations and reduces the number of free parameters of the problem. With this relation we are also able to set constraints to the envelope composition depending on the accretion mass rate. We find that the evolution of MXB 1659-29, KS 1731-260 and EXO 0748-676 can be well described within a deep crustal cooling scenario. Conversely, we find that other two sources can only be explained with models beyond crustal cooling. For the peculiar emission of XTE J1701-462 we propose alternative scenarios like residual accretion during quiescence, additional heat sources in the outer crust and/or thermal isolation of the inner crust due to a buried magnetic field. We also explain the very recent reported temperature of IGR J17480-2446 with an extra heat deposition in the outer crust coming from shallow sources.

This work presents a comparison between the results of experimental tests and Monte Carlo simulations of the efficiency of a detector prototype for on-ground monitoring the cosmic radiation neutron flux. The experimental tests were made using one conventional {sup 241}Am-Be neutron source in several incidence angles and the results were compared to that ones obtained with a Monte Carlo simulation made with MCNPX Code.

A new method of detecting radiation which can allow for long distance measurements is being investigated. The device is primarily for neutrons detection althought it could, in principle, be used for gamma ray detection. The neutron detection medium is a solid, transparent, electro-optical material, such as lithium niobate, lithium tantalite, or barium borate. Crystals of these materials act as optical gates to laser light, allowing light to pass through only when a neutron interaction occurs in the crystal. Typical light detection devices, such as CCD cameras or photomultiplier tubes, can be used to signal when light passes through the crystal. The overall goal of the project is to investigate the feasibility of such devices for the detection of neutron radiation and to quantify their capabilities and limitations.

Thesis. The development of a directional high energy (20 to 160 MeV) neutrondetector which was flown to satellite altitudes (500 km; circular equatorial orbit) in the NASA Orbiting Solar Observatory (OSO-6) in August 1969 is described. Both the angle of incidence and the energy of the neutron are determined by a proton-recoil telescope (Pilot B scintillation plastic) which provides the source for proton-recoils and defines the dE/dX versus E method for particle identification and energy determination. The telescope is embedded in a scintillation plastic guard counter envelope which eliminates the unwanted charged particle background as well as recoil protons (electrons) whose energies and direction do not satisfy neutron (gamma-ray) detection requirements, respectively. Results from a Monte Carlo calculation indicate that the overall average efficiency within an average angular acceptance of about 29 deg (FWHM) is approximately (2.25 plus or minus 0.113) x 10/sup -4/. The inflight calibration procedure, the main frame data bit error analysis, and the method for determining the orientation of the detector axis in the spacecraft spin plane are described. Results indicate a discrepancy in the measured (0.461 x 10/sup -2/ plus or minus 0.254 x 10/sup -2/ n/cm/sup 2/sec) and theoretical (2 to 70 n/cm/ sup 2/sec) neutron flux es which suggests a lack of basic underatanding of mechanisms leading to high energy neutron production at the sun. (auth)

For the fission neutron spectrum measurement, the neutron energy is determined in a time-of-flight experiment by the time difference between the fission event and detection of the neutron. Therefore, the neutron energy resolution is directly determined by the time resolution of both neutron and fission detectors. For the fission detection, the detector needs not only a good timing response but also the tolerance of radiation damage and high {alpha}-decay rate. A parallel-plate avalanche counter (PPAC) has many advantages for the detection of heavy charged particles such as fission fragments. These include fast timing, resistance to radiation damage, and tolerance of high counting rate. A PPAC also can be tuned to be insensitive to particles, which is important for experiments with - emitting actinides. Therefore, a PPAC is an ideal detector for experiments requiring a fast and clean trigger for fission. In the following sections, the description will be given for the design and performance of a new low-mass PPAC for the fission-neutron spectrum measurements at LANL.

CR-39 detectors were exposed to DT neutrons generated by a Thermo Fisher model A290 neutron generator. Afterwards, the etched tracks were examined both optically and by SEM. The purpose of the analysis was to compare the two techniques and to determine whether additional information on track geometry could be obtained by SEM analysis. The use of these techniques to examine triple tracks, diagnostic of ?9.6 MeV neutrons, observed in CR-39 used in Pd/D codeposition experiments will also be discussed.

Because of recent advances in experimental techniques, which improved the accuracies of thermal capture and scattering cross sections by an order of magnitude, a more stringent approach in the evaluation of the thermal constants is developed. In the present approach, the following aspects are introduced: (1) a consistency between thermal cross sections, coherent and incoherent scattering lengths, and neutron resonance parameters is achieved; (2) a consistency between the isotopic and element cross sections is sought; in addition, for each isotope, the requirement that the partial cross sections add up to the total is fulfilled; (3) where possible, charged particle data particularly derived from (d,p) reactions on light and medium weight isotopes are used in locating the positions and strengths of bound levels. Such a procedure is useful in the evaluation of the shape of the cross sections in the thermal region; and (4) the Lane-Lynn theory of direct capture is called upon to calculate thermal cross sections and check for consistencies for certain isotopes. Extensive examples to illustrate these procedures are presented.

One embodiment includes a material exhibiting an optical response signature for neutrons that is different than an optical response signature for gamma rays, said material exhibiting performance comparable to or superior to stilbene in terms of distinguishing neutrons from gamma rays, wherein the material is not stilbene. Another embodiment includes a substantially pure crystal exhibiting an optical response signature for neutrons that is different than an optical response signature for gamma rays, the substantially pure crystal comprising a material selected from a group consisting of: 1-1-4-4-tetraphenyl-1-3-butadiene; 2-fluorobiphenyl-4-carboxylic acid; 4-biphenylcarboxylic acid; 9-10-diphenylanthracene; 9-phenylanthracene; 1-3-5-triphenylbenzene; m-terphenyl; bis-MSB; p-terphenyl; diphenylacetylene; 2-5-diphenyoxazole; 4-benzylbiphenyl; biphenyl; 4-methoxybiphenyl; n-phenylanthranilic acid; and 1-4-diphenyl-1-3-butadiene.

The existing Monte Carlo N-Particle (MCNPX) particle tracking (PTRAC) coincidence capture file allows a full list of neutron capture events to be recorded in any simulated detection medium. The originating event history number (e.g. spontaneous fission events), capture time, location and source particle number are tracked and output to file for post-processing. We have developed a new MCNPX PTRAC coincidence capture file capability to aid detector design studies. New features include the ability to track the nuclides that emitted the detected neutrons as well as induced fission chains in mixed samples before detection (both generation number and nuclide that underwent induced fission). Here, the power of this tool is demonstrated using a detector design developed for the non-destructive assay (NDA) of spent nuclear fuel. Individual capture time distributions have been generated for neutrons originating from Curium-244 source spontaneous fission events and induced fission events in fissile nuclides of interest: namely Plutonium-239, Plutonium-241, and Uranium-235. Through this capability, a full picture for the attribution of neutron capture events in the detector can be simulated.

The existing Monte Carlo N-Particle (MCNPX) particle tracking (PTRAC) coincidence capture file allows a full list of neutron capture events to be recorded in any simulated detection medium. The originating event history number (e.g. spontaneous fission events), capture time, location and source particle number are tracked and output to file for post-processing. We have developed a new MCNPX PTRAC coincidence capture file capability to aid detector design studies. New features include the ability to track the nuclides that emitted the detected neutrons as well as induced fission chains in mixed samples before detection (both generation number and nuclide that underwent induced fission). Here, the power of this tool is demonstrated using a detector design developed for the non-destructive assay (NDA) of spent nuclear fuel. Individual capture time distributions have been generated for neutrons originating from Curium-244 source spontaneous fission events and induced fission events in fissile nuclides of interest: namely Plutonium-239, Plutonium-241, and Uranium-235. Through this capability, a full picture for the attribution of neutron capture events in the detector can be simulated.

The existing MCNPX{trademark} PTRAC coincidence capture file allows a full list of neutron capture events to be recorded in any simulated detection medium. The originating event history number (e.g. spontaneous fission events), capture time, location and source particle number are tracked and output to file for post-processing. We have developed a new MCNPX PTRAC coincidence capture file capability to aid detector design studies. New features include the ability to track the isotopes that emitted the detected neutrons as well as induced fission chains in mixed samples before detection (both generation number and isotope). Here, the power of this tool is demonstrated using a detector design that has been developed for the non-destructive assay (NDA) of spent nuclear fuel. Individual capture time distributions have been generated for neutrons originating from Curium-244 source spontaneous fission events and induced fission events in fissile isotopes of interest: namely Plutonium-239, Plutonium-241, and Uranium-235. Through this capability, a full picture for the attribution of neutron capture events in the detector can be simulated.

The Low Energy NeutronDetector Array (LENDA) is a neutron time-of-flight (TOF) spectrometer developed at the National Superconducting Cyclotron Lab- oratory (NSCL) for use in inverse kinematics experiments with rare isotope beams. Its design has been motivated by the need to study the spin-isospin response of unstable nuclei using (p, n) charge-exchange reactions at intermediate energies (> 100 MeV/u). It can be used, however, for any reaction study that involves emission of low energy neutrons (150 keV - 10 MeV). The array consists of 24 plastic scintillator bars and is capable of registering the recoiling neutron energy and angle with high detection efficiency. The neutron energy is determined by the time-of-flight technique, while the position of interaction is deduced using the timing and energy information from the two photomultipliers of each bar. A simple test setup utilizing radioactive sources has been used to characterize the array. Results of test measurements are compared with simulations. A neutron energy threshold of 20 % for neutrons below 4 MeV have been obtained.

Sample records for thermal neutron detector from the National Library of Energy Beta (NLEBeta)

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The Phase II period performance was May 30, 2002 through May 29, 2004. This development effort was successfully completed within the period and budget allotted. The proposed design was successfully fabricated from B{sub 4}C-coated aluminum and copper film, slit and wound to form 4 mm diameter straws, cut to 100 cm in length, and threaded with resistive anode wires (20 {micro}m in diameter). This paper reports testing done with two 50-straw detector modules at the reactor of the Nuclear Science Center at Texas A&M University (TAMU NSC).

A {chi}{sup 2} minimization algorithm has been developed to extract sub-sampling-time information from digitized waveforms, to be used to instrument the future Versatile Array of NeutronDetectors at Low energies. The algorithm performance has been characterized with a fast Arbitrary Function Generator, obtaining time resolution better than 1 ns for signals of amplitudes between 50 mV and 1V, with negligible walk in the whole range. The proof-of-principle measurement of the beta-delayed neutron emission from {sup 89}Br indicates a resolution of 1 ns can be achieved in realistic experimental conditions.

The National Nuclear Data Center (NNDC) presents two tables showing energy and photon intensity with uncertainties of gamma rays as seen in thermal-neutron capture. One table is organized in ascending order of gamma energy, and the second is organized by Z, A of the target. In the energy-ordered table the three strongest transitions are indicated in each case. The nuclide given is the target nucleus in the capture reaction. The gamma energies given are in keV. The gamma intensities given are relative to 100 for the strongest transition. %I? (per 100 n-captures) for the strongest transition is given, where known. All data are taken from the Evaluated Nuclear Structure Data File (ENSDF), a computer file of evaluated nuclear structure data and from the eXperimental Unevaluated Nuclear Data List (XUNDL). (Specialized Interface)

Reported here are the results of measurements performed to determine the efficiency of 3He filled proportional counters as a function of gas pressure in the SAIC system. Motivation for these measurements was largely to validate the current model of the SAIC system. Those predictions indicated that the neutron detection efficiency plotted as a function of pressure has a simple, logarithmic shape. As for absolute performance, the model results indicated the 3He pressure in the current SAIC system could not be reduced appreciably while meeting the current required level of detection sensitivity. Thus, saving 3He by reducing its pressure was predicted not to be a viable option in the current SAIC system.

This paper investigates the reliability of different noise estimators aimed at determining the Moderator Temperature Coefficient (MTC) of reactivity in Pressurized Water Reactors. By monitoring the inherent fluctuations in the neutron flux and moderator temperature, an on-line monitoring of the MTC without perturbing reactor operation is possible. In order to get an accurate estimation of the MTC by noise analysis, the point-kinetic component of the neutron noise and the core-averaged moderator temperature noise have to be used. Because of the scarcity of the in-core instrumentation, the determination of these quantities is difficult, and several possibilities thus exist for estimating the MTC by noise analysis. Furthermore, the effect of feedback has to be negligible at the frequency chosen for estimating the MTC in order to get a proper determination of the MTC. By using an integrated neutronic/thermal- hydraulic model specifically developed for estimating the three-dimensional distributions of the fluctuations in neutron flux, moderator properties, and fuel temperature, different approaches for estimating the MTC by noise analysis can be tested individually. It is demonstrated that a reliable MTC estimation can only be provided if the core is equipped with a sufficient number of both neutrondetectors and temperature sensors, i.e. if the core contain in-core detectors monitoring both the axial and radial distributions of the fluctuations in neutron flux and moderator temperature. It is further proven that the effect of feedback is negligible for frequencies higher than 0.1 Hz, and thus the MTC noise estimations have to be performed at higher frequencies. (authors)

The thermalneutron absorption cross sections of geologic materials are of first-order importance to the interpretation of pulsed neutron porosity logs and of second-order importance to the interpretation of steady-state porosity logs using dual detectors. Even in the latter case, uncertainties in log response can be excessive whenever formations are encountered that possess absorption properties appreciably greater than the limestones used in most tool calibrations. These effects are of importance to logging operations directed at geothermal applications where formation vary from igneous to sedimentary and which may contain solution-deposited minerals with very large cross-section values. Most measurements of cross-section values for geologic materials have been made for hydrocarbon production applications. Hence, the specimen materials are sedimentary and clean in the sense that they are not altered by geothermal fluids. This investigation was undertaken to measure cross-section values from a sequence of igneous materials obtained from a single hole drilled in an active hydrothermal system. 3 refs., 1 fig.

The impact of strong magnetic fields B>10e13 G on the thermal evolution of neutron stars is investigated, including crustal heating by magnetic field decay. For this purpose, we perform 2D cooling simulations with anisotropic thermal conductivity considering all relevant neutrino emission processes for realistic neutron stars. The standard cooling models of neutron stars are called into question by showing that the magnetic field has relevant (and in many cases dominant) effects on the thermal evolution. The presence of the magnetic field significantly affects the thermal surface distribution and the cooling history of these objects during both, the early neutrino cooling era and the late photon cooling era. The minimal cooling scenario is thus more complex than generally assumed. A consistent magneto-thermal evolution of magnetized neutron stars is needed to explain the observations.

The impact of strong magnetic fields B>10e13 G on the thermal evolution of neutron stars is investigated, including crustal heating by magnetic field decay. For this purpose, we perform 2D cooling simulations with anisotropic thermal conductivity considering all relevant neutrino emission processes for realistic neutron stars. The standard cooling models of neutron stars are called into question by showing that the magnetic field has relevant (and in many cases dominant) effects on the thermal evolution. The presence of the magnetic field significantly affects the thermal surface distribution and the cooling history of these objects during both, the early neutrino cooling era and the late photon cooling era. The minimal cooling scenario is thus more complex than generally assumed. A consistent magneto-thermal evolution of magnetized neutron stars is needed to explain the observations.

Mirror thermal noise is and will remain one of the main limitations to the sensitivity of gravitational wave detectors based on laser interferometers. We report about projected mirror thermal noise due to losses in the mirror coatings and substrates. The evaluation includes all kind of thermal noises presently known. Several of the envisaged substrate and coating materials are considered. The results for mirrors operated at room temperature and at cryogenic temperature are reported.

A process is given for determining the neutronic purity of fissionable material by the so-called shotgun test. The effect of a standard neutron absorber of known characteristics and amounts on a neutronic field also of known characteristics is measured and compared with the effect which the impurities derived from a known quantity of fissionable material has on the same neutronic field. The two readings are then made the basis of calculation from which the amount of impurities can be computed.

Institute of Nuclear Physics the source of epithermal neutrons based on a vacuum insulation tandem At the Budker Institute of Nuclear Physics the VITA-facility for the boron neutron capture therapyNew technical solution for using the time-of-flight technique to measure neutron spectra V. Aleynik

The quest for improved neutron capture cross sections for advanced reactor concepts, transmutation of radioactive wastes as well as for astrophysical scenarios of neutron capture nucleosynthesis has motivated new experimental efforts based on modern techniques. Recent measurements in the keV region have shown that a 4p BaF2 detector represents an accurate and versatile instrument for such studies. The present work deals with the potential of such a 4p BaF2 detector in combination with spallation neutron sources, which offer large neutron fluxes over a wide energy range. Detailed Monte Carlo simulations with the GEANT package have been performed to investigate the critical backgrounds at a spallation facility, to optimize the detector design, and to discuss alternative solutions.

As a promising new fuel for high power density light water reactors, the feasibility of using annular fuel for BWR services is explored from both thermal hydraulic and neutronic points of view. Keeping the bundle size ...

Neutron diffusion length in reactor grade graphite is measured both experimentally and theoretically. The experimental work includes Monte Carlo (MC) coding using 'MCNP' and Finite Element Analysis (FEA) coding suing 'COMSOL Multiphysics' and Matlab. The MCNP code is adopted to simulate the thermalneutron diffusion length in a reactor moderator of 2m x 2m with slightly enriched uranium ($^{235}U$), accompanied with a model designed for thermal hydraulic analysis using point kinetic equations, based on partial and ordinary differential equation. The theoretical work includes numerical approximation methods including transcendental technique to illustrate the iteration process with the FEA method. Finally collision density of thermalneutron in graphite is measured, also specific heat relation dependability of collision density is also calculated theoretically, the thermalneutron diffusion length in graphite is evaluated at $50.85 \\pm 0.3cm$ using COMSOL Multiphysics and $50.95 \\pm 0.5cm$ using MCNP. Finally ...

The quantum properties are important to study nuclear structure. The energy, spin, parity, transition order are usually interesting to research. In this experiment, 59Ni is activated by thermalneutron on 3rd horizontal channel of Dalat nuclear Reactor.

The event-by-event fission model FREYA is extended to spontaneous fission of actinides and a variety of neutron observables are studied for spontaneous fission and fission induced by thermalneutrons with a view towards possible applications for detection of special nuclear materials.

The event-by-event fission model FREYA is extended to spontaneous fission of actinides and a variety of neutron observables are studied for spontaneous fission and fission induced by thermalneutrons with a view towards possible applications for detection of special nuclear materials.

Radiochemically determined mass-yield curves are given for the fission of U/sup 235/ and U/sup 238/ by l4.7-Mev neutrons. Symmetric and to a less extent, very asymmetric modes of fission are more probable at that energy than in thermal fission. Yields of four fission products from the fission of U/sup 235/ have been measured as a function of neutron energy in the range thermal to 14 Mev. The yields of eleven masses have been measured from the fission of Np/sup 237/ by degraded fission spectrum neutrons. The mass-yield curve is similar to that from the thermal fission of Pu/sup 239/ with a ratio of peak to valley yields of approximately 175. Relative yields of one peak fission product and four valley fission products have been determdned under the following conditions: fission of U/sup 235/ and Pu/sup 239/ with thermalneutrons; fission of U/sup 235/ Pu/sup 239/ and U/sup 238/ with fission spectrum neutrons; and fission of U//sup 235/ and Pu/sup 239/ with the intermediate neutron spectrum at the center of the Los Alamos Fast Reactor. Absolute yields of Moss have been determined from the fission of U/sup 235/,Pu/sup 239/ with thermalneutrons. (auth)

A water Cerenkov-based neutron and high energy gamma ray detector and radiation portal monitoring system using water doped with a Gadolinium (Gd)-based compound as the Cerenkov radiator. An optically opaque enclosure is provided surrounding a detection chamber filled with the Cerenkov radiator, and photomultipliers are optically connected to the detect Cerenkov radiation generated by the Cerenkov radiator from incident high energy gamma rays or gamma rays induced by neutron capture on the Gd of incident neutrons from a fission source. The PMT signals are then used to determine time correlations indicative of neutron multiplicity events characteristic of a fission source.

The possibility of using a trap with ultracold neutrons (UCNs) as a detector of dark matter particles with long-range forces is considered. The main advantage of this method is the possibility of detecting recoil energies {approx}10{sup -7} eV. The limitations on the parameters of the interaction potential in the form {Psi}=ae{sup -r/b}/r between dark matter particles and neutrons at different values of the dark matter density on the Earth are represented. It is shown that the suggestion about the long-range character of the interaction between dark matter particles leads to a significant increase in the elastic scattering cross section at low energies. As a consequence, dark matter can be captured and accumulated by the terrestrial gravitational field. The first experimental limitations on the existence of long-range dark matter on the Earth are presented.

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Epithermal neutron data acquired by Mars Odyssey have been analyzed to determine global maps of water-equivalent hydrogen abundance. By assuming that hydrogen was distributed uniformly with depth within the surface, a map of minimum water abundance was obtained. The addition of thermalneutrons to this analysis could provide information needed to determine water stratigraphy. For example, thermal and epithermal neutrons have been used together to determine the depth and abundance of waterequivalent hydrogen of a buried layer in the south polar region. Because the emission of thermalneutrons from the Martian surface is sensitive to absorption by elements other than hydrogen, analysis of stratigraphy requires that the abundance of these elements be known. For example, recently published studies of the south polar region assumed that the Mars Pathfinder mean soil composition is representative of the regional soil composition, This assumption is partially motivated by the fact that Mars appears to have a well-mixed global dust cover and that the Pathfinder soil composition is representative of the mean composition of the Martian surface. In this study, we have analyzed thermal and epithermal neutron data measured by the neutron spectrometer subsystem of the gamma ray spectrometer to determine the spatial distribution of the composition of elements other than hydrogen. We have restricted our analysis to mid-latitude regions for which we have corrected the neutron counting data for variations in atmospheric thickness.

In this paper, we estimate how quickly and how precisely a reactor's operational status and thermal power can be monitored over hour to month time scales, using the antineutrino rate as measured by a cubic meter scale detector. Our results are obtained from a detector we have deployed and operated at 25 meter standoff from a reactor core. This prototype can detect a prompt reactor shutdown within five hours, and monitor relative thermal power to three percent within seven days. Monitoring of short-term power changes in this way may be useful in the context of International Atomic Energy Agency's (IAEA) Reactor Safeguards Regime, or other cooperative monitoring regimes.

In this paper, we estimate how quickly and how precisely a reactor's operational status and thermal power can be monitored over hour to month time scales, using the antineutrino rate as measured by a cubic meter scale detector. Our results are obtained from a detector we have deployed and operated at 25 meter standoff from a reactor core. This prototype can detect a prompt reactor shutdown within five hours, and monitor relative thermal power to three percent within seven days. Monitoring of short-term power changes in this way may be useful in the context of International Atomic Energy Agency's (IAEA) Reactor Safeguards Regime, or other cooperative monitoring regimes.

The thermalneutron cross section and the resonance integral of the reaction {sup 170}Er(n,{gamma}){sup 171}Er were measured by the Cd-ratio method using a {sup 55}Mn monitor as single comparator. Analytical grade MnO{sub 2} and Er{sub 2}O{sub 3} powder samples with and without a cylindrical 1 mm Cd shield box were irradiated in an isotropic neutron field obtained from three {sup 241}Am-Be neutron sources. The induced activities in the samples were measured with a 120.8% relative efficiency p-type HPGe detector. The correction factors for gamma-ray attenuation (F{sub g}), thermalneutron self-shielding (G{sub th}), and resonance neutron self-shielding (G{sub epi}) effects, and the epithermal neutron spectrum shape factor ({alpha}) were taken into account. The thermalneutron cross section for the (n,{gamma}) reaction in {sup 170}Er has been determined to be 8.00 {+-} 0.56 b, relative to that of the {sup 55}Mn monitor. However, some previously reported experimental results compared to the present result show a large discrepancy ranging from 8.3 to 86%. The present result is, in general, in good agreement with the recently measured values by 9%. According to the definition of Cd cut-off energy at 0.55 eV, the resonance integral obtained is 44.5 {+-} 4.0 b, which is determined relative to the reference integral value of the {sup 55}Mn monitor by using cadmium ratios. The existing experimental data for the resonance integral are distributed between 18 and 43 b. The present resonance integral value agrees only with the measurement of 43 {+-} 5 b by Gillette [Thermal Cross Section and Resonance Integral Studies, ORNL-4155, 15 (1967)] within uncertainty limits.

Uncooled pyroelectric IR imaging systems, such as night vision goggles, offer important strategic advantages in battlefield scenarios and reconnaissance surveys. Until now, the current technology for fabricating these devices has been limited by low throughput and high cost which ultimately limit the availability of these sensor devices. We have developed and fabricated an alternative design for pyroelectric IR imaging sensors that utilizes a multilayered thin film deposition scheme to create a monolithic thin film imaging element on an active silicon substrate for the first time. This approach combines a thin film pyroelectric imaging element with a thermally insulating SiO{sub 2} aerogel thin film to produce a new type of uncooled IR sensor that offers significantly higher thermal, spatial, and temporal resolutions at a substantially lower cost per unit. This report describes the deposition, characterization and optimization of the aerogel thermal isolation layer and an appropriate pyroelectric imaging element. It also describes the overall integration of these components along with the appropriate planarization, etch stop, adhesion, electrode, and blacking agent thin film layers into a monolithic structure. 19 refs., 8 figs., 6 tabs.

Over the past year, new 1 m x 1 m neutrondetectors have been installed at both the General Purpose SANS (GP-SANS) and the Bio-SANS instruments at HFIR, each intended as an upgrade to provide improved high rate capability. This paper presents the results of characterization studies performed in the detector test laboratory, including position resolution, linearity and background, as well as a preliminary look at high count rate performance.

A novel method for modeling the neutron time of flight (nTOF) detector response in current mode for inertial confinement fusion experiments has been applied to the on-axis nTOF detectors located in the basement of the Z-Facility. It will be shown that this method can identify sources of neutron scattering, and is useful for predicting detector responses in future experimental configurations, and for identifying potential sources of neutron scattering when experimental set-ups change. This method can also provide insight on how much broadening neutron scattering contributes to the primary signals, which is then subtracted from them. Detector time responses are deconvolved from the signals, allowing a transformation from dN/dt to dN/dE, extracting neutron spectra at each detector location; these spectra are proportional to the absolute yield.

A neutrondetector designed for detecting neutron sources at distances of 50 to 100 m has been constructed and tested. This detector has a large surface area (1 m{sup 2}) to enhance detection efficiency, and it contains a collimator and shielding to achieve direction sensitivity and reduce background. An unusual feature of the detector is that it contains no added moderator, such as polyethylene, to moderate fast neutrons before they reach the {sup 3}He detector. As a result, the detector is sensitive mainly to thermalneutrons. The moderator-free design reduces the weight of the detector, making it more portable, and it also aids in achieving directional sensitivity and background reduction. Test results show that moderated fission-neutron sources of strength about 3 x 10{sup 5} n/s can be detected at a distance out to 70 m in a counting time of 1000 s. The best angular resolution of the detector is obtained at distances of 30 m or less. As the separation .distance between the source and detector increases, the contribution of scattered neutrons to the measured signal increases with a resultant decrease in the ability to detect the direction to a distant source. Applications for which the long-range detector appears to be suitable include detecting remote neutron sources (including sources in moving vehicles) and monitoring neutron storage vaults for the intrusion of humans and the effects they make on the detected neutron signal. Also, the detector can be used to measure waste for the presence of transuranic material in the presence of high gamma-ray background. A test with a neutron source (3 x 10{sup 5} n/s) in a vehicle showed that the detector could readily measure an increase in count rate at a distance of 10 m for vehicle speeds up to 35 mph (the highest speed tested). These results. indicate that the source should be detectable at this distance at speeds up to 55 mph.

Neutron diffusion length in reactor grade graphite is measured both experimentally and theoretically. The experimental work includes Monte Carlo (MC) coding using 'MCNP' and Finite Element Analysis (FEA) coding suing 'COMSOL Multiphysics' and Matlab. The MCNP code is adopted to simulate the thermalneutron diffusion length in a reactor moderator of 2m x 2m with slightly enriched uranium ($^{235}U$), accompanied with a model designed for thermal hydraulic analysis using point kinetic equations, based on partial and ordinary differential equation. The theoretical work includes numerical approximation methods including transcendental technique to illustrate the iteration process with the FEA method. Finally collision density of thermalneutron in graphite is measured, also specific heat relation dependability of collision density is also calculated theoretically, the thermalneutron diffusion length in graphite is evaluated at $50.85 \\pm 0.3cm$ using COMSOL Multiphysics and $50.95 \\pm 0.5cm$ using MCNP. Finally the total neutron cross-section is derived using FEA in an inverse iteration form.

The ATLAS experiment is one of two large general-purpose particle detectors at the Large Hadron Collider (LHC) at the CERN laboratory in Geneva, Switzerland. ATLAS has been collecting data from the collisions of protons since December 2009, in order to investigate the conditions that existed during the early Universe and the origins of mass, and other topics in fundamental particle physics. The innermost layers of the ATLAS detector will be exposed to the most radiation over the first few years of operation at the LHC. In particular, the layer closest to the beam pipe, the B-layer, will degrade over time due to the added radiation. To compensate for its degradation, it will be replaced with an Insertable B-Layer (IBL) around 2016. The design of and R&D for the IBL is ongoing, as the hope is to use the most current technologies in the building of this new sub-detector layer. One topic of interest is the use of more thermally conductive glues in the construction of the IBL, in order to facilitate in the dissipation of heat from the detector. In this paper the measurement and use of highly thermally conductive glues, in particular those that are diamond-loaded, will be discussed. The modified transient plane source technique for thermal conductivity is applied in characterizing the glues across a wide temperature range.

An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented.

Detectors are made of many layers specialized to identify and record Detectors are made of many layers specialized to identify and record information about the many particles that result from a collision of a proton and an antiproton. A sphere would be the best shape to surround the collision point, but it is cheaper to make cylindrical detectors. Because the particles in the Fermilab accelerator have so much energy, detectors may be 3-5 stories high. The layers of a generic detector: (Run the cursor over the names.) Beam Pipe Tracker Electromagnetic Calorimeter Hadron Calorimeter Magnet Muon Detector Anatomy of a Detector - Identifying Particles - CDF Detector - D0 Detector - Links Project Contact: Thomas Jordan - jordant@fnal.gov Web Maintainer: qnet-webmaster@fnal.gov Last Update: April 13, 2001 http://quarknet.fnal.gov/run2/news

We study the participant-spectator matter at the energy of vanishing flow for neutron-rich systems. Our study reveals similar behaviour of articipant-spectator for neutron-rich systems as for stable systems and also points towards nearly mass independence behaviour of participant-spectator matter for neutron-rich systems at the energy of vanishing flow. We also study the thermalization reached in the reactions of neutron-rich systems.

From earlier results on the measurement of soil humidity an apparatus was constructed and calibrated for the measurement of the humidity of soils by diffusion of a beam of thermalneutrons. The construction and calibration of this apparatus are described in detail. (J.S.R.)

An MCNPX-based calculational methodology has been developed to numerically simulate the complex electronphotonneutron transport problem for the active interrogation system known as the pulsed photonuclear assessment (PPA) technique. The PPA technique uses a pulsed electron accelerator to generate bremsstrahlung photons in order to fission nuclear materials. Delayed neutron radiation is then detected with helium-3 neutrondetectors as evidence of the nuclear material presence. Two experimental tests were designed, setup and run to generate experimental data for benchmarking purposes. The first test irradiated depleted uranium in air, and the second test, depleted uranium in a simulated cargo container (plywood pallet), using 10 MeV electron pulses. Time-integrated, post-flash, delayed neutron counts were measured and compared to calculated count predictions in order to benchmark the calculational methodology and computer models. Comparisons between the experimental measurements and numerical predictions of the delayed neutrondetector responses resulted in reasonable experiment/calculated ratios of 1.42 and 1.06 for the two tests. High-enriched uranium (HEU) predictions were also made with the benchmarked models.

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

SABRE (Space-Arena Baseline Reactor) is a 100-kW/sub e/, heat-pipe-cooled, beryllium-reflected, fast reactor that produces heat at a temperature of 1500/sup 0/K and radiatively transmits it to high-temperature thermoelectric (TE) conversion elements. The use of heat pipes for core heat removal eliminates single-point failure mechanisms in the reactor cooling system, and provides minimal temperature drop radiative coupling to the TE array, as well as automatic, self-actuating removal of reactor afterheat. The question of how the failure of a fuel module heat pipe will affect neighboring fuel modules in the core is discussed, as is fission density peaking that occurs at the core/reflector interface. Results of neutronic calculations of the control margin available are described. Another issue that is addressed is that of helium generation in the heat pipes from neutron reactions in the core with the heat pipe fluid. Finally, the growth potential of the SABRE design to much higher powers is examined.

Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermalneutron flux. High thermalneutron fluxes generated from the action of a high power proton accelerator on a spallation target allows the efficient burn-up of higher actinide nuclear waste by a two-step process. Additionally, rapid burn-up of fission product waste for nuclides having small thermalneutron cross sections, and the practicality of small material inventories while achieving significant throughput derive from employment of such high fluxes. Several nuclear technology problems are addressed including 1. nuclear energy production without a waste stream requiring storage on a geological timescale, 2. the burn-up of defense and commercial nuclear waste, and 3. the production of defense nuclear material. The apparatus includes an accelerator, a target for neutron production surrounded by a blanket region for transmutation, a turbine for electric power production, and a chemical processing facility. In all applications, the accelerator power may be generated internally from fission and the waste produced thereby is transmuted internally so that waste management might not be required beyond the human lifespan.

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Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermalneutron flux. High thermalneutron fluxes generated from the action of a high power proton accelerator on a spallation target allows the efficient burn-up of higher actinide nuclear waste by a two-step process. Additionally, rapid burn-up of fission product waste for nuclides having small thermalneutron cross sections, and the practicality of small material inventories while achieving significant throughput derive from employment of such high fluxes. Several nuclear technology problems are addressed including 1. nuclear energy production without a waste stream requiring storage on a geological timescale, 2. the burn-up of defense and commercial nuclear waste, and 3. the production of defense nuclear material. The apparatus includes an accelerator, a target for neutron production surrounded by a blanket region for transmutation, a turbine for electric power production, and a chemical processing facility. In all applications, the accelerator power may be generated internally from fission and the waste produced thereby is transmuted internally so that waste management might not be required beyond the human lifespan.

To measure high-level-activity scrap and waste, it is necessary to use neutrondetectors that are insensitive to the high gamma-ray background. We have developed a combination of {sup 3}He tubes and custom preamplifiers to provide the high efficiency associated with {sup 3}He detectors with good gamma-ray rejection. We have preamplifiers with short time constants in the signal processing to help separate the neutron signals from the slower risetime gamma signals. We have compared AMPTEK (A-111) preamplifiers with Precision Data Technology (PDT 110A) preamplifiers with experimental tests for gamma rejection and radiation damage. Hot cell radiation tests using a 4.5 Ci radium source were performed using {sup 10}B and {sup 3}He detectors to evaluate relative efficiency and the ability to separate neutrons and gamma rays. The AMPTEK A-111 and PDT-110A amplifiers were exposed to gamma doses between {approximately}0.1 R/h and 1500 R/h to observe where the gamma pileup would interfere with the neutron counting. The conclusion is that both amplifiers can operate in gamma fields up to {approximately}500 R/h with modest loss of neutron efficiency. This is valid for the case of only one {sup 3}He tube (30-cm active length) connected to a single amplifier. If an amplifier services multiple tubes or longer tubes, the gamma rejection will get worse. Studies are in progress to determine the lifetime of the amplifiers and {sup 3}He tubes in the high-radiation fields.

A weighted point model for thermalneutron multiplicity counting has been developed for the assay of impure plutonium metal samples. Weighting factors are introduced for the spontaneous fission and ({alpha},n) contributions to the doubles and triples rates to account for the variations in neutron multiplication in these samples. The weighting factors are obtained from Monte Carlo simulations using the MCNPX code, which supports the simulation of spontaneous fission sources and can tally the source and detected neutron multiplicity distributions. Systematic behavior of the weighting factors was studied as a function of sample mass and geometry. Simulations were performed to evaluate the potential accuracy of assays performed with weighted point model analysis. Comparisons with experimental data are presented. The possible use of quads rates is explored.

Since 1967, the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL) has provided state-of-the-art experimental irradiation testing capability. A unique design is investigated herein for the purpose of providing a fast neutron flux test capability in the ATR. This new test capability could be brought on line in approximately 5 or 6 years, much sooner than a new test reactor could be built, to provide an interim fast-flux test capability in the timeframe before a fast-flux research reactor could be built. The proposed cost for this system is approximately $63M, much less than the cost of a new fast-flux test reactor. A concept has been developed to filter out a large portion of the thermal flux component by using a thermally conductive neutron absorber block. The objective of this study is to determine the feasibility of this experiment cooling concept.

We have carried out numerical simulations of the thermal hydraulic behavior of a neutron spallation target where liquid metal lead-bismuth serves as both coolant and as a neutron spallation source. The target is one of three designs provided by the Institute of Physics and Power Engineering (IPPE) in Russia. This type of target is proposed for Accelerator-driven Transmutation of Waste (ATW) to eliminate plutonium from hazardous fission products. The thermal hydraulic behavior was simulated by use of a commercial CFD computer code called CFX. Maximum temperatures in the diaphragm window and in the liquid lead were determined. In addition the total pressure drop through the target was predicted. The results of the CFX analysis were close to those results predicted by IPPE in their preliminary analysis.

Uranium formations are assayed by prompt fission neutron logging techniques. The uranium in the formation is proportional to the ratio of epithermal counts to thermal or eqithermal dieaway. Various calibration factors enhance the accuracy of the measurement.

Uranium formations are assayed by prompt fission neutron logging techniques. The uranium in the formation is proportional to the ratio of epithermal counts to thermal or epithermal dieaway. Various calibration factors enhance the accuracy of the measurement.

During the last few years, the Electric Power Research Institute (EPRI) has led an effort to conduct functionality analyses, which have been performed by AREVA and Westinghouse, to evaluate the potential degradation of reactor internals during life extension to 60 years. The methodologies employed in these functionality analyses make use of different thermal-hydraulic, neutronics, and structural models and computer codes and involve different geometries and loading histories. A comparative ...

A device for accurately measuring the mass of /sup 240/Pu and /sup 239/Pu in a sample having arbitrary moderation and mixed with various contaminants. The device utilizes a thermalneutron well counter which has two concentric rings of neutrondetectors separated by a moderating material surrounding the well. Neutron spectroscopic information derived by the two rings of detectors is used to measure the quantity of /sup 239/Pu and /sup 240/Pu in device which corrects for background radiation, deadtime losses of the detector and electronics and various other constants of the system.

At thermalneutron energies, the binding of the scattering nucleus in a solid, liquid, or gas affects the cross section and the distribution of secondary neutrons. These effects are described in the thermal sub-library of Version VI of the Evaluated Nuclear Data Files (ENDF/B-VI) using the File 7 format. In the original release of the ENDF/B-VI library, the data in File 7 were obtained by converting the thermal scattering evaluations of ENDF/B-III to the ENDF-6 format. These original evaluations were prepared at General Atomics (GA) in the late sixties, and they suffer from accuracy limitations imposed by the computers of the day. This report describes new evaluations for six of the thermal moderator materials and six new cold moderator materials. The calculations were made with the LEAPR module of NJOY, which uses methods based on the British code LEAP, together with the original GA physics models, to obtain new ENDF files that are accurate over a wider range of energy and momentum transfer than the existing files. The new materials are H in H{sub 2}O, Be metal, Be in BeO, C in graphite, H in ZrH, Zr in ZrH, liquid ortho-hydrogen, liquid para-hydrogen, liquid ortho-deuterium, liquid para-deuterium liquid methane, and solid methane.

The average of fragment kinetic energy (E-bar sign*) and the multiplicity of prompt neutrons ({nu}(bar sign)) as a function of fragment mass (m*), as well as the fragment mass yield (Y(m*)) from thermalneutron-induced fission of {sup 239}Pu have been measured by Tsuchiya et al.. In that work the mass and kinetic energy are calculated from the measured kinetic energy of one fragment and the difference of time of flight of the two complementary fragments. However they do not present their results about the standard deviation {sigma}{sub E}*(m*). In this work we have made a numerical simulation of that experiment which reproduces its results, assuming an initial distribution of the primary fragment kinetic energy (E(A)) with a constant value of the standard deviation as function of fragment mass ({sigma}{sub E}(A)). As a result of the simulation we obtain the dependence {sigma}{sub E}*(m*) which presents an enhancement between m* = 92 and m* = 110, and a peak at m* = 121.

The Combined Thermal/Epithermal Neutron (CTEN) non-destructive assay (NDA) system was designed to assay transuranic waste by employing an induced active neutron interrogation and/or a spontaneous passive neutron measurement. This is the second of two papers, and focuses on the passive mode, relating the net double neutron coincidence measurement to the plutonium mass via the calibration constant. National Institute of Standards and Technology (NIST) calibration standards were used and the results verified with NIST-traceable verification standards. Performance demonstration program (PDP) 'empty' 208-L matrix drum was used for the calibration. The experimentally derived calibration constant was found to be 0.0735 {+-} 0.0059 g {sup 240}Pu effective per unit response. Using this calibration constant, the Waste Isolation Pilot Plant (WIPP) criteria was satisfied with five minute waste assays in the range from 3 to 177g Pu. CTEN also participated in the PDP Cycle 8A blind assay with organic sludge and metal matrices and passed the criteria for accuracy and precision in both assay modes. The WIPP and EPA audit was completed March 1, 2002 and full certification is awaiting the closeout of one finding during the audit. With the successful closeout of the audit, the CTEN system will have shown that it can provide very fast assays (five minutes or less) of waste in the range from the minimum detection limit (about 2 mg Pu) to 177 g Pu.

The relative isotopic abundances and the fisson yields for over 40 stable and long-lived fission products from /sup 239/Pu fast fission were evaluated to determine if the data could be correlated with neutron energy. Only mass spectrometric data were used in this study. For some nuclides changes of only a few percent in the relative isotopic abundance or the fission yields over the energy range of thermal to 1 MeV are easily discernable and significant; for others the data are too sparse and scattered to obtain a good correlation. The neutron energy index usedin this study is the /sup 150/Nd//sup 143/Nd isotopic ratio. The results of this correlation study compared to the US Evaluated Nuclear Data File (ENDF) fast fission yield compilation. Several discrepancies are noted and suggestions for future work are presented.

A large heat load caused by thermal radiation through a metal shield pipe was observed in a cooling test of a cryostat for a prototype of a cryogenic interferometric gravitational wave detector. The heat load was approximately 1000 times larger than the value calculated by the Stefan-Boltzmann law. We studied this phenomenon by simulation and experiment and found that it was caused by the conduction of thermal radiation in a metal shield pipe.

A scintillator for neutron time-of-flight measurements is positioned at a desired angle with respect to the neutron beam, and as a function of the energy thereof, such that the sum of the transit times of the neutrons and photons in the scintillator are substantially independent of the points of scintillations within the scintillator. Extrapolated zero timing is employed rather than the usual constant fraction timing. As a result, a substantially larger scintillator can be employed that substantially increases the data rate and shortens the experiment time.

Methods and systems for the detection of small amounts of modern, highly-explosive nitrogen-based explosives, such as plastic explosives, hidden in airline baggage. Several techniques are employed either individually or combined in a hybrid system. One technique employed in combination is X-ray imaging. Another technique is interrogation with a pulsed neutron source in a two-phase mode of operation to image both nitrogen and oxygen densities. Another technique employed in combination is neutron interrogation to form a hydrogen density image or three-dimensional map. In addition, deliberately-placed neutron-absorbing materials can be detected.

Methods and systems for the detection of small amounts of modern, highly-explosive nitrogen-based explosives, such as plastic explosives, hidden in airline baggage. Several techniques are employed either individually or combined in a hybrid system. One technique employed in combination is X-ray imaging. Another technique is interrogation with a pulsed neutron source in a two-phase mode of operation to image both nitrogen and oxygen densities. Another technique employed in combination is neutron interrogation to form a hydrogen density image or three-dimensional map. In addition, deliberately-placed neutron-absorbing materials can be detected.

Sample records for thermal neutron detector from the National Library of Energy Beta (NLEBeta)

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The limitation of current remmeters, which do not measure neutron dose equivalents above about 15 MeV, is a serious problem at high-energy accelerator facilities, where a much wider range of neutron energies exist. The purpose of this work was to measure the response of a modified Anderson-Braun (A-B) remmeter to neutron energies up to 1 GeV. The modifications to the standard A-B remmeter were based on the experimental results of Pb(n,xn) reactions.

At the Karlsruhe pulsed 3.75\\,MV Van de Graaff accelerator the thermonuclear $^{48}$Ca(n,$\\gamma$)$^{49}$Ca(8.72\\,min) cross section was measured by the fast cyclic activation technique via the 3084.5\\,keV $\\gamma$-ray line of the $^{49}$Ca-decay. Samples of CaCO$_3$ enriched in $^{48}$Ca by 77.87\\,\\% were irradiated between two gold foils which served as capture standards. The capture cross-section was measured at the neutron energies 25, 151, 176, and 218\\,keV, respectively. Additionally, the thermal capture cross-section was measured at the reactor BR1 in Mol, Belgium, via the prompt and decay $\\gamma$-ray lines using the same target material. The $^{48}$Ca(n,$\\gamma$)$^{49}$Ca cross-section in the thermonuclear and thermal energy range has been calculated using the direct-capture model combined with folding potentials. The potential strengths are adjusted to the scattering length and the binding energies of the final states in $^{49}$Ca. The small coherent elastic cross section of $^{48}$Ca+n is explained through the nuclear Ramsauer effect. Spectroscopic factors of $^{49}$Ca have been extracted from the thermal capture cross-section with better accuracy than from a recent (d,p) experiment. Within the uncertainties both results are in agreement. The non-resonant thermal and thermonuclear experimental data for this reaction can be reproduced using the direct-capture model. A possible interference with a resonant contribution is discussed. The neutron spectroscopic factors of $^{49}$Ca determined from shell-model calculations are compared with the values extracted from the experimental cross sections for $^{48}$Ca(d,p)$^{49}$Ca and $^{48}$Ca(n,$\\gamma$)$^{49}$Ca.

A novel algorithm for the discrimination of neutron and {\\gamma}-ray with wavelet transform modulus maximum (WTMM) in an organic scintillation has been investigated. Voltage pulses arising from a BC501A organic liquid scintillation detector in a mixed radiation field have been recorded with a fast digital sampling oscilloscope. The performances of most pulse shape discrimination methods in scintillation detection systems using time-domain features of the pulses are affected intensively by noise. However, the WTMM method using frequency-domain features exhibits a strong insensitivity to noise and can be used to discriminate neutron and {\\gamma}-ray events based on their different asymptotic decay trend between the positive modulus maximum curve and the negative modulus maximum curve in the scale-space plane. This technique has been verified by the corresponding mixed-field data assessed by the time-of-flight (TOF) method and the frequency gradient analysis (FGA) method. It is shown that the characterization of neutron and gamma achieved by the discrimination method based on WTMM is consistent with that afforded by TOF and better than FGA. Moreover, because the WTMM method is it self presented to eliminate the noise, there is no need to make any pretreatment for the pulses.

The CERN (European Organization for Nuclear Research) laboratory is currently building the Large Hadron Collider (LHC). Four international collaborations have designed (and are now constructing) detectors able to exploit the physics potential of this collider. Among them is the Compact Muon Solenoid (CMS), a general purpose detector optimized for the search of Higgs boson and for physics beyond the Standard Model of fundamental interactions between elementary particles. This thesis presents, in particular, the design, construction, commissioning and test of the control system for a screen that provides a thermal separation between the Tracker and ECAL (Electromagnetic CALorimeter) detector of CMS (Compact Muon Solenoid experiment). Chapter 1 introduces the new challenges posed by these installations and deals, more in detail, with the Tracker detector of CMS. The size of current experiments for high energy physics is comparable to that of a small industrial plant: therefore, the techniques used for controls a...

This is a report on the CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements held at Annapolis, Maryland, USA November 5-8, 1996. This experts` meeting consisted of 140 participants from 21 countries; 65 invited papers were presented. The meeting was divided into five areas: (1) current and prospective plans of thermal hydraulic codes development; (2) current and anticipated uses of thermal-hydraulic codes; (3) advances in modeling of thermal-hydraulic phenomena and associated additional experimental needs; (4) numerical methods in multi-phase flows; and (5) programming language, code architectures and user interfaces. The workshop consensus identified the following important action items to be addressed by the international community in order to maintain and improve the calculational capability: (a) preserve current code expertise and institutional memory, (b) preserve the ability to use the existing investment in plant transient analysis codes, (c) maintain essential experimental capabilities, (d) develop advanced measurement capabilities to support future code validation work, (e) integrate existing analytical capabilities so as to improve performance and reduce operating costs, (f) exploit the proven advances in code architecture, numerics, graphical user interfaces, and modularization in order to improve code performance and scrutibility, and (g) more effectively utilize user experience in modifying and improving the codes.

The merger of two neutron stars leaves behind a rapidly spinning hypermassive object whose survival is believed to depend on the maximum mass supported by the nuclear equation of state, angular momentum redistribution by (magneto-)rotational instabilities, and spindown by gravitational waves. The high temperatures (~5-40 MeV) prevailing in the merger remnant may provide thermal pressure support that could increase its maximum mass and, thus, its life on a neutrino-cooling timescale. We investigate the role of thermal pressure support in hypermassive merger remnants by computing sequences of spherically-symmetric and axisymmetric uniformly and differentially rotating equilibrium solutions to the general-relativistic stellar structure equations. Using a set of finite-temperature nuclear equations of state, we find that hot maximum-mass critically spinning configurations generally do not support larger baryonic masses than their cold counterparts. However, subcritically spinning configurations with mean density of less than a few times nuclear saturation density yield a significantly thermally enhanced mass. Even without decreasing the maximum mass, cooling and other forms of energy loss can drive the remnant to an unstable state. We infer secular instability by identifying approximate energy turning points in equilibrium sequences of constant baryonic mass parametrized by maximum density. Energy loss carries the remnant along the direction of decreasing gravitational mass and higher density until instability triggers collapse. Since configurations with more thermal pressure support are less compact and thus begin their evolution at a lower maximum density, they remain stable for longer periods after merger.

The high-humidity and high-temperature response of the Eberline Model PRS-2 portable scaler-ratemeter and the Eberline Model NRD neutrondetector was studied in an environmental chamber. The BF/sub 3/ probe used in the NRD detector was found to produce count rate surges at temperatures > 50/sup 0/C and at relative humidity > 50%. The PRS-2 scaler-ratemeter was found to be relatively insensitive to high temperatures and high humidity.

An integrated nondestructive assay instrument, which combined the Passive Neutron Albedo Reactivity (PNAR) and the Self-Interrogation Neutron Resonance Densitometry (SINRD) techniques, is the research focus for a collaborative effort between Los Alamos National Laboratory (LANL) and the Japanese Atomic Energy Agency as part of the Next Generation Safeguard Initiative. We will quantify the anticipated performance of this experimental system in two physical environments: (1) At LANL we will measure fresh Low Enriched Uranium (LEU) assemblies for which the average enrichment can be varied from 0.2% to 3.2% and for which Gd laced rods will be included. (2) At Fugen we will measure spent Mixed Oxide (MOX-B) and LEU spent fuel assemblies from the heavy water moderated Fugen reactor. The MOX-B assemblies will vary in burnup from {approx}3 GWd/tHM to {approx}20 GWd/tHM while the LEU assemblies ({approx}1.9% initial enrichment) will vary from {approx}2 GWd/tHM to {approx}7 GWd/tHM. The estimated count rates will be calculated using MCNPX. These preliminary results will help the finalization of the hardware design and also serve a guide for the experiment. The hardware of the detector is expected to be fabricated in 2012 with measurements expected to take place in 2012 and 2013. This work is supported by the Next Generation Safeguards Initiative, Office of Nuclear Safeguards and Security, National Nuclear Security Administration.

The overall obljectives of this project are to critically review the currently used thermalneutron scattering laws for various moderators as a function of temperature, select as well documented and representative set of experimental data sensitive to the neutron spectra to generate a data base of benchmarks, update models and models parameters by introducing new developments in thermalization theory and condensed matter physics into various computational approaches in establishing the scattering laws, benchmark the results against the experimentatl set. In the case of graphite, a validation experiment is performed by observing nutron slowing down as a function of temperatures equal to or greater than room temperature.

Multiplicity counters for neutron assay have been extensively used in materials control and accountability for nonproliferation and nuclear safeguards. Typically, neutron coincidence counters are utilized in these fields. In this work, we present a measurement system that makes use not only of neutron (n) multiplicity counting but also of gamma-ray (g) multiplicity counting and the combined higher-order multiples containing both neutrons and gamma rays. The benefit of this approach is in using both particle types available from the sample, leading to a reduction in measurement times needed when using more measurables. We present measurement results of n, g, nn, ng, gg, nnn, nng, ngg, and ggg multiples emitted by Mixed-Oxide (MOX) samples measured at Idaho National Laboratory (INL). The MOX measurement is compared to initial validation of the detection system done using a 252Cf source. The dual radiation measuring system proposed here uses extra measurables to improve the statistics when compared to a neutron-only system and allows for extended analysis and interpretation of sample parameters. New challenges such as the effect of very high intrinsic gamma-ray sources in the case of MOX samples is discussed. Successful measurements of multiples rates can be performed also when using high-Z shielding.

An innovative helium3 high pressure gas detection system, made possible by utilizing Sandia's expertise in Micro-electrical Mechanical fluidic systems, is proposed which appears to have many beneficial performance characteristics with regards to making these neutron measurements in the high bremsstrahlung and electrical noise environments found in High Energy Density Physics experiments and especially on the very high noise environment generated on the fast pulsed power experiments performed here at Sandia. This same system may dramatically improve active WMD and contraband detection as well when employed with ultrafast (10-50 ns) pulsed neutron sources.

A conduction cooled neutron absorber may include a metal matrix composite that comprises a metal having a thermalneutron cross-section of at least about 50 barns and a metal having a thermal conductivity of at least about 1 W/cmK. Apparatus for providing a neutron flux having a high fast-to-thermalneutron ratio may include a source of neutrons that produces fast neutrons and thermalneutrons. A neutron absorber positioned adjacent the neutron source absorbs at least some of the thermalneutrons so that a region adjacent the neutron absorber has a fast-to-thermalneutron ratio of at least about 15. A coolant in thermal contact with the neutron absorber removes heat from the neutron absorber.

The mass and kinetic energy distribution of nuclear fragments from thermalneutron-induced fission of {sup 235}U(n{sub th},f) have been studied using a Monte-Carlo simulation. Besides reproducing the pronounced broadening in the standard deviation of the kinetic energy at the final fragment mass number around m = 109, our simulation also produces a second broadening around m = 125. These results are in good agreement with the experimental data obtained by Belhafaf et al. and other results on yield of mass. We conclude that the obtained results are a consequence of the characteristics of the neutron emission, the sharp variation in the primary fragment kinetic energy and mass yield curves. We show that because neutron emission is hazardous to make any conclusion on primary quantities distribution of fragments from experimental results on final quantities distributions.

Ternary fission probabilities for thermalneutron induced fission of plutonium are analyzed within the framework of an evaporation-based model where the complexity of time-varying potentials, associated with the neck collapse, are included in a simplistic fashion. If the nuclear temperature at scission and the fission-neck-collapse time are assumed to be ~1.2 MeV and ~10^-22 s, respectively, then calculated relative probabilities of ternary-fission light-charged-particle emission follow the trends seen in the experimental data. The ability of this model to reproduce ternary fission probabilities spanning seven orders of magnitude for a wide range of light-particle charges and masses implies that ternary fission is caused by the coupling of an evaporation-like process with the rapid re-arrangement of the nuclear fluid following scission.

, 110 8th Street, Troy, NY USA 12180-3522 ABSTRACT Detection of nuclear materials is critical-be terrorist also increases. With these issues in mind, research has begun in the area of making a low cost to accurately transport and track alpha particle and light ion energy deposition in the detector. The following

In a passive multiplicity characterization of highly enriched uranium (HEU) assemblies, fission chains are initiated by the characteristically fast neutrons from spontaneous fission of {sup 238}U and {sup 235}U as well as cosmic-ray spallation neutrons. Active interrogation of HEU uses other physical mechanisms for starting chains by inducing fission from high-energy neutrons, high-energy gamma-rays, delayed neutrons, or thermalneutrons. In all cases a contribution to the initiation of fission chains is the reflection of neutrons that initially escape the assembly and re-enter it after undergoing some scattering. The reflected neutron flux is geometry dependent and a combination of fast and thermal energies. The reflected thermalneutron contribution occurs hundreds of microseconds after the beginning of the fission chain and can be distinguished from the cosmic-ray spallation neutrons unrelated to fission chains, resulting in an HEU detection signature with high signal-to-noise. However, the reflected thermalneutron flux can be eliminated with an efficient thermalneutron absorber to investigate reflected neutron effects. In this paper, active and passive multiplicity measurements with HEU oxide assemblies of up to 16 kg of fuel pins and HEU metal assemblies of up to five 18 kg storage castings are reported. Each case demonstrates the differences in HEU signature when a borated thermalneutron absorber is present and shows the various detectable signatures with 3He proportional counters, the standard detector for differential die-way and neutron multiplicity measurements, and liquid scintillators, a detector capable of operating on the timescale of fission chains.

For many years at LLNL we have been developing time-correlated neutron detection techniques and algorithms for many applications including Arms Control, Threat Detection and Nuclear Material Assaying. Many of our techniques have been developed specifically for relatively low efficiency (a few %) inherent in the man-portable systems. Historically we used thermalneutrondetectors (mainly {sup 3}He) taking advantage of the high thermalneutron interaction cross-sections but more recently we have been investigating fast neutron detection with liquid scintillators and inorganic crystals. We have discovered considerable detection advantages with fast neutron detection as the inherent nano-second production time-scales of fission and neutron induced fission are preserved instead of being lost in neutronthermalization required for thermalneutrondetectors. We are now applying fast neutron technology (new fast and portable digital electronics as well as new faster and less hazardous scintillator formulations) to the safeguards regime and faster detector response times and neutron momentum sensitivity show promise in measuring, differentiating and assaying samples that have very high count rates as well as mixed fission sources (e.g. Cm and Pu). We report on measured results with our existing liquid scintillator array and progress on design of nuclear material assaying system that incorporates fast neutron detection.

The relatively large number of nearby radio-quiet and thermally emitting isolated neutron stars (INSs) discovered in the ROSAT All-Sky Survey, dubbed the ``Magnificent Seven'' (M7), suggests that they belong to a formerly neglected major component of the overall INS population. So far, attempts to discover similar INSs beyond the solar vicinity failed to confirm any reliable candidate. The EPIC cameras onboard the XMM-Newton satellite allow to efficiently search for new thermally emitting INSs. We used the 2XMMp catalogue to select sources with no catalogued candidate counterparts and with X-ray spectra similar to those of the M7, but seen at greater distances and thus undergoing higher interstellar absorptions. Identifications in more than 170 astronomical catalogues and visual screening allowed to select fewer than 30 good INS candidates. In order to rule out alternative identifications, we obtained deep ESO-VLT and SOAR optical imaging for the X-ray brightest candidates. We report here on the optical follo...

The structure of deuterated jarosite, KFe{sub 3}(SO{sub 4}){sub 2}(OD){sub 6}, was investigated using time-of-flight neutron diffraction up to its dehydroxylation temperature. Rietveld analysis reveals that with increasing temperature, its c dimension expands at a rate {approx}10 times greater than that for a. This anisotropy of thermal expansion is due to rapid increase in the thickness of the (001) sheet of [Fe(O,OH){sub 6}] octahedra and [SO{sub 4}] tetrahedra with increasing temperature. Fitting of the measured cell volumes yields a coefficient of thermal expansion, a = a{sub 0} + a{sub 1} T, where a{sub 0} = 1.01 x 10{sup -4} K{sup -1} and a{sub 1} = -1.15 x 10{sup -7} K{sup -2}. On heating, the hydrogen bonds, O1{hor_ellipsis}D-O3, through which the (001) octahedral-tetrahedral sheets are held together, become weakened, as reflected by an increase in the D{hor_ellipsis}O1 distance and a concomitant decrease in the O3-D distance with increasing temperature. On further heating to 575 K, jarosite starts to decompose into nanocrystalline yavapaiite and hematite (as well as water vapor), a direct result of the breaking of the hydrogen bonds that hold the jarosite structure together.

Natural gadolinium is used as a burnable poison in most LWR to account for the excess of reactivity of fresh fuels. For an accurate prediction of the cycle length, its nuclear data and especially its neutron capture cross section needs to be known with a high precision. Recent microscopic measurements at Rensselaer Polytechnic Inst. (RPI) suggest a 11% smaller value for the thermal capture cross section of {sup 157}Gd, compared with most of evaluated nuclear data libraries. To solve this inconsistency, we have analyzed several pile-oscillation experiments, performed in the MINERVE reactor. They consist in the measurement of the reactivity variation involved by the introduction in the reactor of small-samples, containing different mass amounts of natural gadolinium. The analysis of these experiments is done through the exact perturbation theory, using the PIMS calculation tool, in order to link the reactivity effect to the thermal capture cross section. The measurement of reactivity effects is used to deduce the 2200 m.s-1 capture cross section of {sup nat}Gd which is (49360 {+-} 790) b. This result is in good agreement with the JEFF3.1.1 value (48630 b), within 1.6% uncertainty at 1{sigma}, but is strongly inconsistent with the microscopic measurements at RPI which give (44200 {+-} 500) b. (authors)

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There are several machines in this country that produce short bursts of neutrons for various applications. A few examples are the Zmachine, operated by Sandia National Laboratories in Albuquerque, NM; the OMEGA Laser Facility at the University of Rochester in Rochester, NY; and the National Ignition Facility (NIF) operated by the Department of Energy at Lawrence Livermore National Laboratory in Livermore, California. They all incorporate neutron time of flight (nTOF) detectors which measure neutron yield, and the shapes of the waveforms from these detectors contain germane information about the plasma conditions that produce the neutrons. However, the signals can also be %E2%80%9Cclouded%E2%80%9D by a certain fraction of neutrons that scatter off structural components and also arrive at the detectors, thereby making analysis of the plasma conditions more difficult. These detectors operate in current mode - i.e., they have no discrimination, and all the photomultiplier anode charges are integrated rather than counted individually as they are in single event counting. Up to now, there has not been a method for modeling an nTOF detector operating in current mode. MCNPPoliMiwas developed in 2002 to simulate neutron and gammaray detection in a plastic scintillator, which produces a collision data output table about each neutron and photon interaction occurring within the scintillator; however, the postprocessing code which accompanies MCNPPoliMi assumes a detector operating in singleevent counting mode and not current mode. Therefore, the idea for this work had been born: could a new postprocessing code be written to simulate an nTOF detector operating in current mode? And if so, could this process be used to address such issues as the impact of neutron scattering on the primary signal? Also, could it possibly even identify sources of scattering (i.e., structural materials) that could be removed or modified to produce %E2%80%9Ccleaner%E2%80%9D neutron signals? This process was first developed and then applied to the axial neutron time of flight detectors at the ZFacility mentioned above. First, MCNPPoliMi was used to model relevant portions of the facility between the source and the detector locations. To obtain useful statistics, variance reduction was utilized. Then, the resulting collision output table produced by MCNPPoliMi was further analyzed by a MATLAB postprocessing code. This converted the energy deposited by neutron and photon interactions in the plastic scintillator (i.e., nTOF detector) into light output, in units of MeVee%D1%84 (electron equivalent) vs time. The time response of the detector was then folded into the signal via another MATLAB code. The simulated response was then compared with experimental data and shown to be in good agreement. To address the issue of neutron scattering, an %E2%80%9CIdeal Case,%E2%80%9D (i.e., a plastic scintillator was placed at the same distance from the source for each detector location) with no structural components in the problem. This was done to produce as %E2%80%9Cpure%E2%80%9D a neutron signal as possible. The simulated waveform from this %E2%80%9CIdeal Case%E2%80%9D was then compared with the simulated data from the %E2%80%9CFull Scale%E2%80%9D geometry (i.e., the detector at the same location, but with all the structural materials now included). The %E2%80%9CIdeal Case%E2%80%9D was subtracted from the %E2%80%9CFull Scale%E2%80%9D geometry case, and this was determined to be the contribution due to scattering. The time response was deconvolved out of the empirical data, and the contribution due to scattering was then subtracted out of it. A transformation was then made from dN/dt to dN/dE to obtain neutron spectra at two different detector locations.

There are several machines in this country that produce short bursts of neutrons for various applications. A few examples are the Zmachine, operated by Sandia National Laboratories in Albuquerque, NM; the OMEGA Laser Facility at the University of Rochester in Rochester, NY; and the National Ignition Facility (NIF) operated by the Department of Energy at Lawrence Livermore National Laboratory in Livermore, California. They all incorporate neutron time of flight (nTOF) detectors which measure neutron yield, and the shapes of the waveforms from these detectors contain germane information about the plasma conditions that produce the neutrons. However, the signals can also be %E2%80%9Cclouded%E2%80%9D by a certain fraction of neutrons that scatter off structural components and also arrive at the detectors, thereby making analysis of the plasma conditions more difficult. These detectors operate in current mode - i.e., they have no discrimination, and all the photomultiplier anode charges are integrated rather than counted individually as they are in single event counting. Up to now, there has not been a method for modeling an nTOF detector operating in current mode. MCNPPoliMiwas developed in 2002 to simulate neutron and gammaray detection in a plastic scintillator, which produces a collision data output table about each neutron and photon interaction occurring within the scintillator; however, the postprocessing code which accompanies MCNPPoliMi assumes a detector operating in singleevent counting mode and not current mode. Therefore, the idea for this work had been born: could a new postprocessing code be written to simulate an nTOF detector operating in current mode? And if so, could this process be used to address such issues as the impact of neutron scattering on the primary signal? Also, could it possibly even identify sources of scattering (i.e., structural materials) that could be removed or modified to produce %E2%80%9Ccleaner%E2%80%9D neutron signals? This process was first developed and then applied to the axial neutron time of flight detectors at the ZFacility mentioned above. First, MCNPPoliMi was used to model relevant portions of the facility between the source and the detector locations. To obtain useful statistics, variance reduction was utilized. Then, the resulting collision output table produced by MCNPPoliMi was further analyzed by a MATLAB postprocessing code. This converted the energy deposited by neutron and photon interactions in the plastic scintillator (i.e., nTOF detector) into light output, in units of MeVee%D1%84 (electron equivalent) vs time. The time response of the detector was then folded into the signal via another MATLAB code. The simulated response was then compared with experimental data and shown to be in good agreement. To address the issue of neutron scattering, an %E2%80%9CIdeal Case,%E2%80%9D (i.e., a plastic scintillator was placed at the same distance from the source for each detector location) with no structural components in the problem. This was done to produce as %E2%80%9Cpure%E2%80%9D a neutron signal as possible. The simulated waveform from this %E2%80%9CIdeal Case%E2%80%9D was then compared with the simulated data from the %E2%80%9CFull Scale%E2%80%9D geometry (i.e., the detector at the same location, but with all the structural materials now included). The %E2%80%9CIdeal Case%E2%80%9D was subtracted from the %E2%80%9CFull Scale%E2%80%9D geometry case, and this was determined to be the contribution due to scattering. The time response was deconvolved out of the empirical data, and the contribution due to scattering was then subtracted out of it. A transformation was then made from dN/dt to dN/dE to obtain neutron spectra at two different detector locations.

Multi-scale, multi-physics problems reveal significant challenges while dealing with coupled neutronic/thermal-hydraulic solutions. Current generation of codes applied to Light Water Reactors (LWR) are based on 3D neutronic nodal methods coupled with one or two phase flow thermal-hydraulic system or sub-channel codes. In addition, spatial meshing and temporal schemes are crucial for the proper description of the non-symmetrical core behavior in case of transient and accidents e.g. reactivity insertion accidents. This paper describes the coupling approach between the 3D neutron diffusion code COBAYA3 and the sub-channel code SUBCHANFLOW within SALOME. The coupling is done inside the SALOME open source platform that is characterized by a powerful pre- and post-processing capabilities and a novel functionality for mapping of the neutronic and thermal hydraulic domains. The peculiar functionalities of SALOME and the steps required for the code integration and coupling are presented. The validation of the coupled codes is done based on two benchmarks the PWR MOX/UO{sub 2} RIA and the TMI-1 MSLB benchmark. A discussion of the prediction capability of COBAYA3/SUBCHANFLOW compared to other coupled solutions will be provided too. (authors)

Neutron Measurements / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection

For many years at LLNL, we have been developing time-correlated neutron detection techniques and algorithms for applications such as Arms Control, Threat Detection and Nuclear Material Assay. Many of our techniques have been developed specifically for the relatively low efficiency (a few percent) attainable by detector systems limited to man-portability. Historically, we used thermalneutrondetectors (mainly {sup 3}He), taking advantage of the high thermalneutron interaction cross-sections. More recently, we have been investigating the use of fast neutron detection with liquid scintillators, inorganic crystals, and in the near future, pulse-shape discriminating plastics which respond over 1000 times faster (nanoseconds versus tens of microseconds) than thermalneutrondetectors. Fast neutron detection offers considerable advantages, since the inherent nanosecond production time-scales of spontaneous fission and neutron-induced fission are preserved and measured instead of being lost by thermalization required for thermalneutrondetectors. We are now applying fast neutron technology to the safeguards regime in the form of fast portable digital electronics as well as faster and less hazardous scintillator formulations. Faster detector response times and sensitivity to neutron momentum show promise for measuring, differentiating, and assaying samples that have modest to very high count rates, as well as mixed fission sources like Cm and Pu. We report on measured results with our existing liquid scintillator array, and progress on the design of a nuclear material assay system that incorporates fast neutron detection, including the surprising result that fast liquid scintillator detectors become competitive and even surpass the precision of {sup 3}He-based counters measuring correlated pairs in modest (kg) samples of plutonium.

We have incorporated neutron-absorbing elements in transparent, non-scintillating glasses and used the Cherenkov effect to convert neutron-induced beta-gamma radiation directly into light. Use of the Cherenkov effect requires glasses with a high index of refraction (to lower the threshold and increase the number of Cherenkov photons), and neutron absorbers resulting in radioactive products emitting high-energy beta or gamma radiation. In this paper, we present a brief description of the requirements for developing efficient Cherenkov-based neutrondetectors, show the results of measurements of the response of representative samples to a thermalneutron flux, and give the results of a calculation of the expected response of a detector to a moderated fission spectrum.

Thermal aging of three-wire series-arc stainless steel weld overlay cladding at 288{degrees}C for 1605 h resulted in an appreciable decrease (16%) in the Charpy V-notch (CVN) upper-shelf energy (USE), but the effect on the 41-J transition temperature shift was very small (3{degrees}C). The combined effect of aging and neutron irradiation at 288{degrees}C to a fluence of 5 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV) was a 22% reduction in the USE and a 29{degrees}C shift in the 41-J transition temperature. The effect of thermal aging on tensile properties was very small. However, the combined effect of irradiation and aging was an increase in the yield strength (6 to 34% at test temperatures from 288 to {minus}125{degrees}C) but no apparent change in ultimate tensile strength or total elongation. Neutron irradiation reduced the initiation fracture toughness (J{sub Ic}) much more than did thermal aging alone. Irradiation slightly decreased the tearing modulus, but no reduction was caused by thermal aging alone. Other results from tensile, CVN, and fracture toughness specimens showed that the effects of thermal aging at 288 or 343{degrees}C for 20,000 h each were very small and similar to those at 288{degrees}C for 1605 h. The effects of long-term thermal exposure time (50,000 h and greater) at 288{degrees}C will be investigated as the specimens become available in 1996 and beyond.

The present paper presents the {sup 36}Cl measurement effort in the US. A large number of {sup 36}Cl measurements have been made in both granite and concrete samples obtained from various locations and distances in Hiroshima and Nagasaki. These measurements employed accelerator mass spectrometry (AMS) to quantify the number of atoms of {sup 36}Cl per atom of total Cl in the sample. Results from these measurements are presented here and discussed in the context of the DS02 dosimetry reevaluation effort for Hiroshima and Nagasaki atomic-bomb survivors. The production of {sup 36}Cl by bomb neutrons in mineral samples from Hiroshima and Nagasaki was primarily via the reaction {sup 35}Cl(n,{gamma}){sup 36}Cl. This reaction has a substantial thermalneutron cross-section (43.6 b at 0.025 eV) and the product has a long half-life (301,000 y). hence, it is well suited for neutron-activation detection in Hiroshima and Nagasaki using AMS more than 50 years after the bombings. A less important reaction for bomb neutrons, {sup 39}K(n,{alpha}){sup 36}Cl, typically produces less than 10% of the {sup 36}Cl in mineral samples such as granite and concrete, which contain {approx} 2% potassium. In 1988, only a year after the publication of the DS86 final report (Roesch 1987), it was demonstrated experimentally that {sup 36}Cl measured using AMS should be able to detect the thermalneutron fluences at the large distances most relevant to the A-bomb survivor dosimetry. Subsequent measurements in mineral samples from both Hiroshima and Nagasaki validated the experimental findings. The potential utility of {sup 36}Cl as a thermalneutrondetector in Hiroshima was first presented by Haberstock et al. who employed the Munich AMS facility to measure {sup 36}Cl/Cl ratios in a gravestone from near the hypocenter. That work subsequently resulted in an expanded {sup 36}Cl effort in Germany that paralleled the US work. More recently, there have also been {sup 36}Cl measurements made by a Japanese group. The impetus for the extensive {sup 36}Cl and other neutron activation measurements was the recognized need to validate the neutron component of the dose in Hiroshima. Although this was suggested at the time of the DS86 Final Report, where it was stated that the calculated neutron doses for survivors could possibly be wrong, the paucity of neutron validation measurements available at that time prevented adequate resolution of this matter. It was not until additional measurements and data evaluations were made that it became clear that more work was required to better understand the discrepancies observed for thermalneutrons in Hiroshima. This resulted in a large number of additional neutron activation measurements in Hiroshima and Nagasaki by scientists in the US, Japan, and Germany. The results presented here for {sup 36}Cl, together with measurements made by other scientists and for other isotopes, now provide a much improved measurement basis for the validation of neutrons in Hiroshima.

Recent advancements in the ultra-wide band Radio Frequency Identification (RFID) technology and solid state pillar type neutrondetectors have enabled us to move forward in combining both technologies for advanced neutron monitoring. The LLNL RFID tag is totally passive and will operate indefinitely without the need for batteries. The tag is compact, can be directly mounted on metal, and has high performance in dense and cluttered environments. The LLNL coin-sized pillar solid state neutrondetector has achieved a thermalneutron detection efficiency of 20% and neutron/gamma discrimination of 1E5. These performance values are comparable to a fieldable {sup 3}He based detector. In this paper we will discuss features about the two technologies and some potential applications for the advanced safeguarding of nuclear materials.

Abstract. Two different CERMET fueled Nuclear Thermal Propulsion reactors were modeled to determine the optimum coolant channel surface area to volume ratio required to cool a 25,000 lbf rocket engine operating at a specific impulse of 940 seconds. Both reactor concepts were computationally fueled with hexagonal cross section fuel elements having a flat-to-flat distance of 3.51 cm and containing 60 vol.% UO2 enriched to 93wt.%U235 and 40 vol.% tungsten. Coolant channel configuration consisted of a 37 coolant channel fuel element and a 61 coolant channel model representing 0.3 and 0.6 surface area to volume ratios respectively. The energy deposition from decelerating fission products and scattered neutrons and photons was determined using the MCNP monte carlo code and then imported into the STAR-CCM+ computational fluid dynamics code. The 37 coolant channel case was shown to be insufficient in cooling the core to a peak temperature of 3000 K; however, the 61 coolant channel model shows promise for maintaining a peak core temperature of 3000 K, with no more refinements to the surface area to volume ratio. The core was modeled to have a power density of 9.34 GW/m3 with a thrust to weight ratio of 5.7.

Thermal properties of low-density neutron matter are investigated by determinantal quantum Monte Carlo lattice calculations on 3+1 dimensional cubic lattices. Nuclear effective field theory (EFT) is applied using the pionless single- and two-parameter neutron-neutron interactions, determined from the $^1S_0$ scattering length and effective range. The determination of the interactions and the calculations of neutron matter are carried out consistently by applying EFT power counting rules. The thermodynamic limit is taken by the method of finite-size scaling, and the continuum limit is examined in the vanishing lattice filling limit. The $^1S_0$ pairing gap at $T \\approx 0$ is computed directly from the off-diagonal long-range order of the spin pair-pair correlation function, and is found to be approximately 30% smaller than BCS calculations with the conventional nucleon-nucleon potentials. The critical temperature $T_c$ of the normal-to-superfluid phase transition and the pairing temperature scale $T^\\ast$ are determined, and the temperature-density phase diagram is constructed. The physics of low-density neutron matter is clearly identified as being a BCS-Bose-Einstein condensation crossover.

Conversion of photons into axions under the presence of a strong magnetic field can dim the radiation from magnetized astrophysical objects. Here we perform a detailed calculation aimed at quantifying the signatures of photon-axion conversion in the spectra, light curves, and polarization of neutron stars (NSs). We take into account the energy and angle dependence of the conversion probability and the surface thermal emission from NSs. The latter is computed from magnetized atmosphere models that include the effect of photon polarization mode conversion due to vacuum polarization. The resulting spectral models, inclusive of the general-relativistic effects of gravitational redshift and light deflection, allow us to make realistic predictions for the effects of photon to axion conversion on observed NS spectra, light curves, and polarization signals. We identify unique signatures of the conversion, such as an increase of the effective area of a hot spot as it rotates away from the observer line of sight. For a star emitting from the entire surface, the conversion produces apparent radii that are either larger or smaller (depending on axion mass and coupling strength) than the limits set by NS equations of state. For an emission region that is observed phase-on, photon-axion conversion results in an inversion of the plane of polarization with respect to the no-conversion case. While the quantitative details of the features that we identify depend on NS properties (magnetic field strength and temperature) and axion parameters, the spectral and polarization signatures induced by photon-axion conversion are distinctive enough to make NSs very interesting and promising probes of axion physics.

The number of neutrons emitted by individual fragments from U/sup 235/ fission by thermalneutrons was measured using a large detector filled with a liquid organic cadmiumcontaining scintillator. The numbers of prompt neutrons were measured under 4 pi geometry conditions as a function of fragment mass. The excitation energy spent on prompt neutrons was derived on the basis of Weizsacker's semiempirical formula. A sharp asymmetry was noticed in the distribution of excitation energies between heavy and light fragments. The new data do not agree with the Fong statistical fission theory. (tr-auth)

The yields of more than fifteen fission products have been carefully measured using radiochemical techniques, for {sup 235}U(n,f), {sup 239}Pu(n,f) in a thermal spectrum, for {sup 233}U(n,f), {sup 235}U(n,f), and {sup 239}Pu(n,f) reactions in a fission neutron spectrum, and for {sup 233}U(n,f), {sup 235}U(n,f), {sup 238}U(n,f), and {sup 239}Pu(n,f) for 14.7 MeV monoenergetic neutrons. Irradiations were performed at the EL3 reactor, at the Caliban and Prospero critical assemblies, and at the Lancelot electrostatic accelerator in CEA-Valduc. Fissions were counted in thin deposits using fission ionization chambers. The number of fission products of each species were measured by gamma spectrometry of co-located thick deposits.

A brief discussion of the principles and techniques of chemical analysis by neutron capture gamma radiation is presented, and the widely scattered literature is collected into a single table arranged by element measured.

Nineteen elements were determined in four different grain size fractions of a bulk geological material from Cerro Impacto for a study of the physical (mechanical) concentration process of different elements based upon the hardness of the different minerals. The analysis was performed by excitation of the sample with a high, slow neutron flux followed by gamma-ray spectroscopy with both a conventional Ge(Li) high-energy detector and a low-energy photon detector (LEPD). The accuracy of this method was studied with the use of two standard reference materials, SY-2 and SY-3, which are similar to the real samples. The values determined were also compared with a secondary target x-ray fluorescence method for all the elements that were suitable to both methods. Actually, the x-ray fluorescence method was found to be more complementary than competitive. 10 refs., 2 figs., 4 tabs.

A method of detecting an activator, the method including impinging a receptor material that is not predominately water and lacks a photoluminescent material with an activator and generating Cherenkov effect light due to the activator impinging the receptor material. The method further including identifying a characteristic of the activator based on the light.

We study the influence of nano-scale layers of converters made from natural gadolinium and its 157 isotope into the total efficiency of registration of thermalneutrons. Our estimations show that contribution of low-energy Auger electrons with the runs about nanometers in gadolinium, to the total efficiency of neutron converters in this case is essential and results in growth of the total efficiency of converters. The received results are in good consent to the experimental data.

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We discuss the fission barrier height of neutron-rich nuclei in a r-process site at highly excited state, which is resulted from the beta-decay or the neutron-capture processes. We particularly investigate the sensitivity of the fission barrier height to the temperature, including the effect of pairing phase transition from superfluid to normal fluid phases. To this end, we use the finite-temperature Skyrme-Hartree-Fock-Bogolubov method with a zero-range pairing interaction. We also discuss the temperature dependence of the fission decay rate.

A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

The key fact about fissile material is that a sufficient quantity of the material can produce chains of fissions, including some very long chains. A chain of fissions will give rise to a detected burst of neutrons with longer chains generally producing larger bursts. These bursts produce distinctive time correlations in a detector near the multiplying material. These correlations are measurable and can be analyzed to infer attributes of the fissile material including fissile material mass, assembly neutron multiplication, characteristic fast fission chain evolution time scale, also known as the {alpha} time scale, thermalization time scale. The correlation signal is very robust with respect to background and to neutron absorbing material.

The Self-Powered NeutronDetector (SPND) Measuring System is evaluated to determine its ability to indicate temperatures of the fuel rods in the TMI-2 reactor core during the accident. It is concluded for the following reasons that the SPND Measuring System did not provide fuel rod temperatures during the accident: the heat transfer characteristics vary over a range of five octaves; within the range of 1200 to 1800/sup 0/F, the SPND responds to temperature from convection radiation from the fuel rods and self-heating from the gamma flux; within the range of 1200 to 1800/sup 0/F, the signal cable introduces masking signals that are a function of gamma heating, integrated temperature over the cable, and core water level velocity; the data system's worst-case signal-to-noise ratio from aliasing is 0dB; and the recorder system's worst-case signal-to-noise ratio from aliasing is -24dB.

Neutron spectrometry can play an important role in the detection and identification of neutron-emitting sources in various security applications. In the present work, a portable filtered array neutron spectrometer, consisting of twelve 6LiF-based thermalneutrondetectors embedded within a single heterogeneous volume was designed and its expected performance compared to that of a commercially available 12-sphere Bonner spheres spectrometer. Each detector within the volume was designed to optimally respond to a unique portion of the neutron spectrum by varying the type and thickness of materials used to filter the spectrum as well as the thickness of the moderator in front of the detector. The available design space was permuted and performance metrics developed to identify the optimal geometries. The top performing detector geometries were then combinatorially explored to identify the best array of geometries that yielded the most information about a neutron spectrum. The best performing filtered array was found to provide as much spectral information as, or more than, the commercially available 12-sphere Bonner spheres spectrometer to which it was compared.

The plasma panel sensor (PPS) is a gaseous micropattern radiation detector under current development. It has many operational and fabrication principles common to plasma display panels (PDPs). It comprises a dense matrix of small, gas plasma discharge cells within a hermetically sealed panel. As in PDPs, it uses non-reactive, intrinsically radiation-hard materials such as glass substrates, refractory metal electrodes, and mostly inert gas mixtures. We are developing these devices primarily as thin, low-mass detectors with gas gaps from a few hundred microns to a few millimeters. The PPS is a high gain, inherently digital device with the potential for fast response times, fine position resolution (< 50 m RMS) and low cost. In this paper we report here on prototype PPS experimental results in detecting betas, protons and cosmic muons, and we extrapolate on the PPS potential for applications including detection of alphas, heavy-ions at low to medium energy, thermalneutrons and X-rays.

This presentation represents an overview of the experimental evaluation of a boron-lined proportional technology performed within an NA-241 sponsored project on testing of boron-lined proportional counters for the purpose of replacement of {sup 3}He technologies. The presented boron-lined technology will be utilized in a design of a full scale safeguards neutron coincidence counter. The design considerations and the Monte Carlo performance predictions for the counter are also presented.

Neutron Science Neutron Science Neutron Scattering Science Neutrons are one of the fundamental particles that make up matter and have properties that make them ideal for certain types of research. In the universe, neutrons are abundant, making up more than half of all visible matter. Neutron scattering provides information about the positions, motions, and magnetic properties of solids. When a beam of neutrons is aimed at a sample, many neutrons will pass through the material. But some will interact directly with atomic nuclei and "bounce" away at an angle, like colliding balls in a game of pool. This behavior is called neutron diffraction, or neutron scattering. Using detectors, scientists can count scattered neutrons, measure their energies and the angles at which they scatter, and map their final position

In this paper, the distribution of mass and kinetic energy in the cold region of the thermalneutron induced fission of U 233, U 235 and Pu 239, respectively, is interpreted in terms of nucleon pair-breaking and the Coulomb interaction energy between complementary fragments (Coulomb effect). In order to avoid the erosive consequences of neutron emission, one studies the cold fission regions, corresponding to total kinetic energy (TE) close to the maximum available energy of the reaction (Q). Contrary to expected, in cold fission is not observed high odd-even effect in mass number distribution. Nevertheless, the measured values are compatible with higher odd-even effects on proton or neutron number distribution, respectively. In addition, in cold fission, the minimal total excitation energy (X) is correlated with the Coulomb energy excess, which is defined as the difference between C (the electrostatic interaction energy between complementary fragments taken as spherical in scission configuration) and Q. These...

This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the codes versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implement a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research projects primary objective is to advance the state of the art for reactor analysis.

We describe measurements of fission product data at Los Alamos that are important for determining the number of fissions that have occurred when neutrons are incident on plutonium and uranium isotopes. The fission-spectrum measurements were made using a fission chamber designed by the National Institute for Standards and Technology (NIST) in the BIG TEN critical assembly, as part of the Inter-laboratory Liquid Metal Fast Breeder Reactor (LMFBR) Reaction Rate (ILRR) collaboration. The thermal measurements were made at Los Alamos' Omega West Reactor. A related set of measurements were made of fission-product ratios (so-called R-values) in neutron environments provided by a number of Los Alamos critical assemblies that range from having average energies causing fission of 400-600 keV (BIG TEN and the outer regions of the Flattop-25 assembly) to higher energies (1.4-1.9 MeV) in the Jezebel, and in the central regions of the Flattop-25 and Flattop-Pu, critical assemblies. From these data we determine ratios of fission product yields in different fuel and neutron environments (Q-values) and fission product yields in fission spectrum neutron environments for {sup 99}Mo, {sup 95}Zr, {sup 137}Cs, {sup 140}Ba, {sup 141,143}Ce, and {sup 147}Nd. Modest incident-energy dependence exists for the {sup 147}Nd fission product yield; this is discussed in the context of models for fission that include thermal and dynamical effects. The fission product data agree with measurements by Maeck and other authors using mass-spectrometry methods, and with the ILRR collaboration results that used gamma spectroscopy for quantifying fission products. We note that the measurements also contradict earlier 1950s historical Los Alamos estimates by {approx}5-7%, most likely owing to self-shielding corrections not made in the early thermal measurements. Our experimental results provide a confirmation of the England-Rider ENDF/B-VI evaluated fission-spectrum fission product yields that were carried over to the ENDF/B-VII.0 library, except for {sup 99}Mo where the present results are about 4%-relative higher for neutrons incident on {sup 239}Pu and {sup 235}U. Additionally, our results illustrate the importance of representing the incident energy dependence of fission product yields over the fast neutron energy range for high-accuracy work, for example the {sup 147}Nd from neutron reactions on plutonium. An upgrade to the ENDF library, for ENDF/B-VII.1, based on these and other data, is described in a companion paper to this work.

Apparatus for improved sensitivity and time resolution of a neutron measurement. The detector is provided with an electrode assembly having a neutron sensitive cathode which emits relatively low energy secondary electrons. The neutron sensitive cathode has a large surface area which provides increased sensitivity by intercepting a greater number of neutrons. The cathode is also curved to compensate for differences in transit time of the neutrons emanating from the point source. The slower speeds of the secondary electrons emitted from a certain portion of the cathode are matched to the transit times of the neutrons impinging thereupon.

A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermalneutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermalneutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

In July of 1999 Louisiana State University (LSU) was awarded a two year research grant by the D.O.E. NEER program to develop a methodology for neutron transport calculations using pointwise (PW) nuclear data in the thermal energy range, and to implement the method into the CENTRM transport code being developed at LSU for Oak Ridge National Laboratory (ORNL). This work has extended CENTRM's current epithermal PW calculation to encompass the thermal range, providing a continuous-energy deterministic transport code that can address problems that may not be adequately treated using multigroup methods. The new version of the CENTRM code was completed, and provided to ORNL for inclusion in the next release of the SCALE code system. The new thermal calculation developed by the NEER project is a significant improvement in the CENTRM capability, and should have an impact on criticality and shipping cask analysis done by numerous organizations who use this code system.

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We have completed the design and testing of neutron scattering instrument detectors for powder diffractometers and single crystal diffractometers. These detectors meet the performance requirements for these types of instruments at the Department of Energy Spallation Neutron Source facility.

1. The combination with a plurality of parallel horizontal members arranged in horizontal and vertical rows, the spacing of the members in all horizontal rows being equal throughout, the spacing of the members in all vertical rows being equal throughout; of a shield for a nuclear reactor comprising two layers of rectangular blocks through which the members pass generally perpendicularly to the layers, each block in each layer having for one of the members an opening equally spaced from vertical sides of the block and located closer to the top of the block than the bottom thereof, whereby gravity tends to make each block rotate about the associated member to a position in which the vertical sides of the block are truly vertical, the openings in all the blocks of one layer having one equal spacing from the tops of the blocks, the openings in all the blocks of the other layer having one equal spacing from the tops of the blocks, which spacing is different from the corresponding spacing in the said one layer, all the blocks of both layers having the same vertical dimension or length, the blocks of both layers consisting of relatively wide blocks and relatively narrow blocks, all the narrow blocks having the same horizontal dimension or width which is less than the horizontal dimension or width of the wide blocks, which is the same throughout, each layer consisting of vertical rows of narrow blocks and wide blocks alternating with one another, each vertical row of narrow blocks of each layer being covered by a vertical row of wide blocks of the other layer which wide blocks receive the same vertical row of members as the said each vertical row of narrow blocks, whereby the rectangular perimeters of each block of each layer is completely out of register with that of each block in the other layer.

The plasma panel sensor (PPS) is a gaseous micropattern radiation detector under current development. It has many operational and fabrication principles common to plasma display panels. It comprises a dense matrix of small, gas plasma discharge cells within a hermetically sealed panel. As in plasma display panels, it uses nonreactive, intrinsically radiation-hard materials such as glass substrates, refractory metal electrodes, and mostly inert gas mixtures. We are developing these devices primarily as thin, low-mass detectors with gas gaps from a few hundred microns to a few millimeters. The PPS is a high gain, inherently digital device with the potential for fast response times, fine position resolution (cost. In this paper, we report on prototype PPS experimental results in detecting betas, protons, and cosmic muons, and we extrapolate on the PPS potential for applications including the detection of alphas, heavy ions at low-to-medium energy, thermalneutrons, and X-rays.

In this paper, the distribution of mass and kinetic energy in the cold region of the thermalneutron induced fission of U 233, U 235 and Pu 239, respectively, is interpreted in terms of nucleon pair-breaking and the Coulomb interaction energy between complementary fragments (Coulomb effect). In order to avoid the erosive consequences of neutron emission, one studies the cold fission regions, corresponding to total kinetic energy (TE) close to the maximum available energy of the reaction (Q). Contrary to expected, in cold fission is not observed high odd-even effect in mass number distribution. Nevertheless, the measured values are compatible with higher odd-even effects on proton or neutron number distribution, respectively. In addition, in cold fission, the minimal total excitation energy (X) is correlated with the Coulomb energy excess, which is defined as the difference between C (the electrostatic interaction energy between complementary fragments taken as spherical in scission configuration) and Q. These Coulomb effects increase with the asymmetry of the charge fragmentations. In sum, the experimental data on cold fission suggest that scission configurations explore all the possibilities permitted by the available energy for fission.

This paper describes an unattended mode neutron measurement that can provide the enrichment of the uranium in UF{sub 6} cylinders. The new passive neutron measurement provides better penetration into the uranium mass than prior gamma-ray enrichment measurement methods. The Passive Neutron Enrichment Monitor (PNEM) provides a new measurement technique that uses passive neutron totals and coincidence counting together with neutron self-interrogation to measure the enrichment in the cylinders. The measurement uses the neutron rates from two detector pods. One of the pods has a bare polyethylene surface next to the cylinder and the other polyethylene surface is covered with Cd to prevent thermalneutrons from returning to the cylinder. The primary neutron source from the enriched UF{sub 6} is the alpha-particle decay from the {sub 234}U that interacts with the fluorine to produce random neutrons. The singles neutron counting rate is dominated by the {sub 234}U neutrons with a minor contribution from the induced fissions in the {sub 235}U. However, the doubles counting rate comes primarily from the induced fissions (i.e., multiplication) in the {sub 235}U in enriched uranium. The PNEM concept makes use of the passive neutrons that are initially produced from the {sub 234}U reactions that track the {sub 235}U enrichment during the enrichment process. The induced fission reactions from the thermal-neutron albedo are all from the {sub 235}U and provide a measurement of the {sub 235}U. The Cd ratio has the desirable feature that all of the thermal-neutron-induced fissions in {sub 235}U are independent of the original neutron source. Thus, the ratio is independent of the uranium age, purity, and prior reactor history.

This report documents work performed by Idaho National Laboratory and the University of Michigan in fiscal year (FY) 2012 to examine design parameters related to the use of fast-neutron multiplicity counting for assaying plutonium for materials protection, accountancy, and control purposes. This project seeks to develop a new type of neutron-measurement-based plutonium assay instrument suited for assaying advanced fuel cycle materials. Some current-concept advanced fuels contain high concentrations of plutonium; some of these concept fuels also contain other fissionable actinides besides plutonium. Because of these attributes the neutron emission rates of these new fuels may be much higher, and more difficult to interpret, than measurements made of plutonium-only materials. Fast neutron multiplicity analysis is one approach for assaying these advanced nuclear fuels. Studies have been performed to assess the conceptual performance capabilities of a fast-neutron multiplicity counter for assaying plutonium. Comparisons have been made to evaluate the potential improvements and benefits of fast-neutron multiplicity analyses versus traditional thermal-neutron counting systems. Fast-neutron instrumentation, using for example an array of liquid scintillators such as EJ-309, have the potential to either a) significantly reduce assay measurement times versus traditional approaches, for comparable measurement precision values, b) significantly improve assay precision values, for measurement durations comparable to current-generation technology, or c) moderating improve both measurement precision and measurement durations versus current-generation technology. Using the MCNPX-PoliMi Monte Carlo simulation code, studies have been performed to assess the doubles-detection efficiency for a variety of counter layouts of cylindrical liquid scintillator detector cells over one, two, and three rows. Ignoring other considerations, the best detector design is the one with the most detecting volume. However, operational limitations guide a) the maximum acceptable size of each detector cell (due to PSD performance and maximum-acceptable per-channel data throughput rates, limited by pulse pile-up and the processing rate of the electronics components of the system) and b) the affordability of a system due to the number of total channels of data to be collected and processed. As a first estimate, it appears that a system comprised of two rows of detectors 5" Ű ? 3" would yield a working prototype system with excellent performance capabilities for assaying Pu-containing items and capable of handling high signal rates likely when measuring items with Pu and other actinides. However, it is still likely that gamma-ray shielding will be needed to reduce the total signal rate in the detectors. As a first step prior to working with these larger-sized detectors, it may be practical to perform scoping studies using small detectors, such as already-on-hand 3" Ű ? 3" detectors.

Recent advances in photoconductive and bolometric semiconductor detectors for wavelength 1 mm > {lambda} > 50 {mu}m are reviewed. Progress in detector performance in this photon energy range has been stimulated by new and stringent requirements for ground based, high altitude and space-borne telescopes for astronomical and astrophysical observations. The paper consists of chapters dealing with the various types of detectors: Be and Ga doped Ge photoconductors, stressed Ge:Ga devices and neutron transmutation doped Ge thermistors. Advances in the understanding of basic detector physics and the introduction of modern semiconductor device technology have led to predictable and reliable fabrication techniques. Integration of detectors into functional arrays has become feasible and is vigorously pursued by groups worldwide.

This article presents the experimental work performed in the area of neutrondetector development at the Remote Sensing LaboratoryAndrews Operations (RSL-AO) sponsored by the U.S. Department of Energy, National Nuclear Security Administration (NNSA) in the last four years. During the 1950s neutrondetectors were developed mostly to characterize nuclear reactors where the neutron flux is high. Due to the indirect nature of neutron detection via interaction with other particles, neutron counting and neutron energy measurements have never been as precise as gamma-ray counting measurements and gamma-ray spectroscopy. This indirect nature is intrinsic to all neutron measurement endeavors (except perhaps for neutron spin-related experiments, viz. neutron spin-echo measurements where one obtains ?eV energy resolution). In emergency response situations generally the count rates are low, and neutrons may be scattered around in inhomogeneous intervening materials. It is also true that neutron sensors are most efficient for the lowest energy neutrons, so it is not as easy to detect and count energetic neutrons. Most of the emergency response neutrondetectors are offshoots of nuclear device diagnostics tools and special nuclear materials characterization equipment, because that is what is available commercially. These instruments mostly are laboratory equipment, and not field-deployable gear suited for mobile teams. Our goal is to design and prototype field-deployable, ruggedized, lightweight, efficient neutrondetectors.

A semiconductor detector for ionizing electromagnetic radiation, neutrons, and energetic charged particles. The detecting element is comprised of a compound having the composition I-III-VI.sub.2 or II-IV-V.sub.2 where the "I" component is from column 1A or 1B of the periodic table, the "II" component is from column 2B, the "III" component is from column 3A, the "IV" component is from column 4A, the "V" component is from column 5A, and the "VI" component is from column 6A. The detecting element detects ionizing radiation by generating a signal proportional to the energy deposited in the element, and detects neutrons by virtue of the ionizing radiation emitted by one or more of the constituent materials subsequent to capture. The detector may contain more than one neutron-sensitive component.

A detector apparatus for differentiating between gamma and neutron radiation is provided. The detector includes a pair of differentially shielded Geiger-Mueller tubes. The first tube is wrapped in silver foil and the second tube is wrapped in lead foil. Both the silver and lead foils allow the passage of gamma rays at a constant rate in a gamma ray only field. When neutrons are present, however, the silver activates and emits beta radiation that is also detected by the silver wrapped Geiger-Mueller tube while the radiation detected by the lead wrapped Geiger-Mueller tube remains constant. The amount of radiation impinging on the separate Geiger-Mueller tubes is then correlated in order to distinguish between the neutron and gamma radiations.

The disclosure relates to a battery operated neutron spectrometer/dosimeter utilizing a microprocessor, a built-in tissue equivalent LET neutrondetector, and a 128-channel pulse height analyzer with integral liquid crystal display. The apparatus calculates doses and dose rates from neutrons incident on the detector and displays a spectrum of rad or rem as a function of keV per micron of equivalent tissue and also calculates and displays accumulated dose in millirads and millirem as well as neutron dose rates in millirads per hour and millirem per hour.

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An improved method for delivering thermalneutrons to a subsurface cancer or tumor which has been first doped with a dopant having a high cross section for neutron capture. The improvement is the use of a guide tube in cooperation with a capillary neutron focusing apparatus, or neutron focusing lens, for directing neutrons to the tumor, and thereby avoiding damage to surrounding tissue.

The feasibility of measuring the transuranic (TRU) nuclide content of equipment removed from Hanford`s high-level radioactive-waste tanks has been established for components heavier than about 30 kg/m (20 lbs/ft). This conclusion has been reached based on experience with the TRU assay of waste burial boxes, planned improvements to the assay equipment design and assay methodology, and experimental investigation of neutrondetector performance in high gamma-ray fields. The experiments indicate that the neutrondetectors presently used with Pacific Northwest Laboratory`s box scanner perform correctly in gamma-ray exposure rates of at least 3 R/h. The design of equipment proposed for measuring TRU content incorporates multiple, BF{sub 3}-gas-filled neutron counters in a configuration that is approximately 0.5 m wide and 2 m long, with polyethylene to moderate high-energy neutrons down to thermal energy. Specially developed electrical systems are used to eliminate response to gamma-rays. Performance of the assay would require 10 to 14 hours of time during which close-range access is provided to the waste and its burial container. A standard neutron source, will be placed within the burial container (before inserting components) to allow calibration of the detector. Final calculation of the TRU contamination will utilize plausible conservative assumptions concerning the spatial, isotopic, and elemental distributions of any TRU present. For long-length equipment, the detector array collects data at various positions along the length of the equipment. Separate monitoring of the cosmic-ray-induced neutron background during the assay period will provide confidence that observed changes in counts at the equipment are not related to changing background. Background measurements using the burial container and equipment {open_quotes}skid{close_quotes} will allow compensation for neutrons that are created by cosmic-ray spallation within the burial container.

The objective of this work has been to improve the knowledge of the thermal cross sections of the fissile nuclei as a step toward providing a standard data base for the nuclear industry. The methodology uses a form of the Adler-Adler multilevel-fission theory and Breit-Wigner multilevel-scattering theory. It incorporates these theories in a general nonlinear least-squares (LSQ) fitting program SIGLEARNThe analysis methodology in this work was applied to the thermal data on /sup 235/U. A reference data file has been developed which includes most of the known data of interest. The first important result of this work is the assessment of the shape uncertainties of the partial cross sections. The results of our studies lead to the following values and error estimates for /sup 235/U g factors in a thermal (20.44/sup 0/C) energy spectrum: g/sub f/ = 0.97751 (+-0.11%); g/sub ..gamma../ = 0.98230 (+-0.14%). A second important result of this study is the development of a recommended set of 2200 m/s (0.0253 eV) values of the parameters and the probable range of further adjustment which might be made. The analysis also provides the result of a common interpretation of energy-dependent absolute cross-section data of different measurements to yield a consistent set of experimental 0.0253 eV values with rigorous error estimates. It also provides normalization factors for relative fission and capture cross sections on a common basis with rigorous error estimates. The results of these analyses provide a basis for deciding what new measurements would be most beneficial. The most important of these would be improved direct capture data in the thermal region.

The PCS (Protein Crystallography Station) at Los Alamos Neutron Science Center (LANSCE) is a unique facility in the USA that is designed and optimized for detecting and collecting neutron diffraction data from macromolecular crystals. PCS utilizes the 20 Hz spallation neutron source at LANSCE to enable time-of-flight measurements using 0.6-7.0 {angstrom} neutrons. This increases the neutron flux on the sample by using a wavelength range that is optimal for studying macromolecular crystal structures. The diagram below show a schematic of PCS and photos of the detector and instrument cave.

The Analytic Hierarchy Process (AHP) has been used to help determine the importance of components and phenomena in thermal-hydraulic safety analyses of nuclear reactors. The AHP results are based, in part on expert opinion. Therefore, it is prudent to evaluate the uncertainty of the AHP ranks of importance. Prior applications have addressed uncertainty with experimental data comparisons and bounding sensitivity calculations. These methods work well when a sufficient experimental data base exists to justify the comparisons. However, in the case of limited or no experimental data the size of the uncertainty is normally made conservatively large. Accordingly, the author has taken another approach, that of performing a statistically based uncertainty analysis. The new work is based on prior evaluations of the importance of components and phenomena in the thermal-hydraulic safety analysis of the Advanced Neutron Source Reactor (ANSR), a new facility now in the design phase. The uncertainty during large break loss of coolant, and decay heat removal scenarios is estimated by assigning a probability distribution function (pdf) to the potential error in the initial expert estimates of pair-wise importance between the components. Using a Monte Carlo sampling technique, the error pdfs are propagated through the AHP software solutions to determine a pdf of uncertainty in the system wide importance of each component. To enhance the generality of the results, study of one other problem having different number of elements is reported, as are the effects of a larger assumed pdf error in the expert ranks. Validation of the Monte Carlo sample size and repeatability are also documented.

We have demonstrated the ability to measure the neutron flux produced by the cosmic-ray interaction with nuclei in the ground surface using aerial neutron detection. High energy cosmic-rays (primarily muons with GeV energies) interact with the nuclei in the ground surface and produce energetic neutrons via spallation. At the air-surface interface, the neutrons produced by spallation will either scatter within the surface material, become thermalized and reabsorbed, or be emitted into the air. The mean free path of energetic neutrons in air can be hundreds of feet as opposed to a few feet in dense materials. As such, the flux of neutrons escaping into the air provides a measure of the surface nuclei composition. It has been demonstrated that this effect can be measured at long range using neutrondetectors on low flying helicopters. Radiological survey measurements conducted at Government Wash in Las Vegas, Nevada, have shown that the neutron background from the cosmic-soil interactions is repeatable and directly correlated to the geological data. Government Wash has a very unique geology, spanning a wide variety of nuclide mixtures and formations. The results of the preliminary measurements are presented.

We have demonstrated the ability to measure the neutron flux produced by the cosmic-ray interaction with nuclei in the ground surface using aerial neutron detection. High energy cosmic-rays (primarily muons with GeV energies) interact with the nuclei in the ground surface and produce energetic neutrons via spallation. At the air-surface interface, the neutrons produced by spallation will either scatter within the surface material, become thermalized and reabsorbed, or be emitted into the air. The mean free path of energetic neutrons in air can be hundreds of feet as opposed to a few feet in dense materials. As such, the flux of neutrons escaping into the air provides a measure of the surface nuclei composition. It has been demonstrated that this effect can be measured at long range using neutrondetectors on low flying helicopters. Radiological survey measurements conducted at Government Wash in Las Vegas, Nevada, have shown that the neutron background from the cosmic-soil interactions is repeatable and directly correlated to the geological data. Government Wash has a very unique geology, spanning a wide variety of nuclide mixtures and formations. The results of the preliminary measurements are presented.

In this paper a new type of passive neutrondetector based on the already existing one, CR39, is described. Its operation was verified by three different neutron sources: an Americium-Beryllium (Am241-Be) source; a TRIGA type nuclear reactor; and a fast neutron reactor called TAPIRO. The obtained results, reported here, positively confirm its operation and the accountability of the new developed detecting technique.

Neutron sensors capable of real-time measurement of thermal flux, fast flux, and temperature in a single miniaturized probe are needed in irradiation tests required to demonstrate the performance of candidate new fuels, and cladding materials. In-core ceramic-based miniature neutrondetectors or Micro-Pocket Fission Detectors (MPFDs) have been studied at Kansas State University (KSU). The first MPFD prototypes were tested in various neutron fields at the KSU TRIGA research reactor with successful results. Currently, a United States Department of Energy-sponsored joint KSU/Idaho National Laboratory (INL) effort is underway to develop a high-temperature, high-pressure version of the MPFD using radiation-resistant, high temperature materials, which would be capable of withstanding irradiation test conditions in high performance material and test reactors (MTRs). Ultimately, this more compact, more accurate, and longer lifetime flux sensor for critical mock-ups, existing and advanced reactor designs, high performance MTRs, and transient test reactors has the potential to lead to higher accuracy and resolution data from irradiation testing, more detailed core flux measurements and enhanced fuel assembly processing. Prior evaluations by KSU indicate that these sensors could also be used to monitor burn-up of nuclear fuel. If integrated into nuclear fuel assemblies, MPFDs offer several advantages to current spent fuel management systems.

A benchmark test of the Monte Carlo neutron and photon transport code system (MCNP) was performed using a bare- and energy-moderated {sup 252}Cf fission neutron source which was obtained by transmission through 10-cm-thick iron. An iron plate was used to simulate the effect of the Hiroshima atomic bomb casing. This test includes the activation of indium and nickel for fast neutrons and gold, europium, and cobalt for thermal and epithermal neutrons, which were inserted in the moderators. The latter two activations are also to validate {sup 152}Eu and {sup 60}Co activity data obtained from the atomic bomb-exposed specimens collected at Hiroshima and Nagasaki, Japan. The neutron moderators used were Lucite and Nylon 6 and the total thickness of each moderator was 60 cm or 65 cm. Measured activity data (reaction yield) of the neutron-irradiated detectors in these moderators decreased to about 1/1,000th or 1/10,000th, which corresponds to about 1,500 m ground distance from the hypocenter in Hiroshima. For all of the indium, nickel, and gold activity data, the measured and calculated values agreed within 25%, and the corresponding values for europium and cobalt were within 40%. From this study, the MCNP code was found to be accurate enough for the bare- and energy-moderated {sup 252}Cf neutron activation calculations of these elements using moderators containing hydrogen, carbon, nitrogen, and oxygen. 18 refs., 10 figs., 4 tabs.

The thermal annealing response of several materials as indicated by Charpy transition temperature (TT) and upper-shelf energy (USE), crack initiation toughness, K{sub Jc}, predictive models, and automated-ball indentation (ABI) testing are compared. The materials investigated are representative reactor pressure vessel (RPV) steels (several welds and a plate) that were irradiated for other tasks of the Heavy-Section Steel Irradiation (HSSI) Program and are relatively well characterized in the unirradiated and irradiated conditions. They have been annealed at two temperatures, 343 and 454 C (650 and 850 F) for varying lengths of time. The correlation of the Charpy response and the fracture toughness, ABI, and the response predicted by the annealing model of Eason et al. for these conditions and materials appears to be reasonable. The USE after annealing at the temperature of 454 C appears to recover at a faster rate than the TT, and even over-recovers (i.e., the recovered USE exceeds that of the unirradiated material).

A low efficiency, 2-Dimensional Position Sensitive NeutronDetector based on delay line position encoding is developed. It is designed to handle beam flux of 10{sup 6}-10{sup 7} n/cm{sup 2}/s and for monitoring intensity profiles of neutron beams. The present detector can be mounted in transmission mode, as the hardware allows maximum neutron transmission in sensitive region. Position resolution of 1.2 mm in X and Y directions, is obtained. Online monitoring of beam images and intensity profile of various neutron scattering spectrometers at Dhruva are presented. It shows better dynamic range of intensity over commercial neutron camera and is also time effective over the traditionally used photographic method.

The delayed neutron emission rates of U-235 and Pu-239 samples were measured accurately from a thermal fission reaction. A Monte Carlo calculation using the Geant4 code was used to demonstrate the neutron energy independence of the detector used in the counting station. A set of highly purified actinide samples (U-235 and Pu-239) was irradiated in these experiments by using the Texas A&M University Nuclear Science Center Reactor. A fast pneumatic transfer system, an integrated computer control system, and a graphite-moderated counting system were constructed to perform all these experiments. The calculated values for the five-group U-235 delayed neutron parameters and the six-group Pu-239 delayed neutron parameters were compared with the values recommended by Keepin et al. (1957) and Waldo et al. (1981). These new values differ slightly from literature values. The graphite-moderated counting station and the computerized pneumatic system are now operational for further delayed neutron measurement.

The first several campaigns of laser fusion experiments at the National Ignition Facility (NIF) included a family of high-sensitivity scintillator/photodetector neutron-time-of-flight (nTOF) detectors for measuring DD and DT neutron yields. The detectors provided consistent neutron yield benchmarks from below 1E9 (DD) to nearly 1E15 (DT). The detectors demonstrated DT yield measurement precisions better than 5%, but the absolute accuracy relies on cross calibration with independent measurements of absolute neutron yield. The 4.5-m nTOF data have provided a useful testbed for testing improvements in nTOF data processing, especially with respect to improving the accuracies of the detector impulse response functions. The resulting improvements in data analysis methods have produced more accurate results. In summary, results from the NIF 4.5-m nTOF detectors have provided consistent measurements of DD and DT neutron yields from laser-fusion implosions.

A radiation detector of the type is described wherein a condenser is directly connected to the electrodes for the purpose of performing the dual function of a guard ring and to provide capacitance coupling for resetting the detector system.

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Neutron, gamma and charged particle detection improvements are key to supporting many of the foreseen measurements and systems envisioned in the R&D programs and the future fuel cycle requirements, such as basic nuclear physics and data, modeling and simulation, reactor instrumentation, criticality safety, materials management and safeguards. This task will focus on the developmental needs of the FCR&D experimental programs, such as elastic/inelastic scattering, total cross sections and fission neutron spectra measurements, and will leverage a number of existing neutrondetector development efforts and programs, such as those at LANL, PNNL, INL, and IAC as well as those at many universities, some of whom are funded under NE grants and contracts. Novel materials and fabrication processes combined with state-of-the-art electronics and computing provide new opportunities for revolutionary detector systems that will be able to meet the high precision needs of the program. This work will be closely coordinated with the Nuclear Data Crosscut. The Advanced Detector Development effort is a broadly-focused activity that supports the development of improved nuclear data measurements and improved detection of nuclear reactions and reactor conditions. This work supports the design and construction of large-scale, multiple component detectors to provide nuclear reaction data of unprecedented quality and precision. Examples include the Time Projection Chamber (TPC) and the DANCE detector at LANL. This work also supports the fabrication and end-user application of novel scintillator materials detection and monitoring.

Passive neutron multiplicity counting is commonly used to quantify the total mass of plutonium in a sample, without prior knowledge of the sample geometry. However, passive neutron counting is less applicable to uranium measurements due to the low spontaneous fission rates of uranium. Active neutron multiplicity measurements are therefore used to determine the {sup 235}U mass in a sample. Unfortunately, there are still additional challenges to overcome for uranium measurements, such as the coupling of the active source and the uranium sample. Techniques, such as the coupling method, have been developed to help reduce the dependence of calibration curves for active measurements on uranium samples; although, they still require similar geometry known standards. An advanced active neutron multiplicity measurement method is being developed by Texas A&M University, in collaboration with Los Alamos National Laboratory (LANL) in an attempt to overcome the calibration curve requirements. This method can be used to quantify the {sup 235}U mass in a sample containing uranium without using calibration curves. Furthermore, this method is based on existing detectors and nondestructive assay (NDA) systems, such as the LANL Epithermal Neutron Multiplicity Counter (ENMC). This method uses an inexpensive boron carbide liner to shield the uranium sample from thermal and epithermal neutrons while allowing fast neutrons to reach the sample. Due to the relatively low and constant fission and absorption energy dependent cross-sections at high neutron energies for uranium isotopes, fast neutrons can penetrate the sample without significant attenuation. Fast neutron interrogation therefore creates a homogeneous fission rate in the sample, allowing for first principle methods to be used to determine the {sup 235}U mass in the sample. This paper discusses the measurement method concept and development, including measurements and simulations performed to date, as well as the potential limitations.

A neutron coincidence module was designed using multistage shift registers to produce the coincidence gates and a crystal controlled oscillator with variable clock outputs to change the gate lengths. The advantage of this system over the conventional, thermal-neutron coincidence gates is a decrease in deadtime by more than an order of magnitude. (auth)

Fissionable uranium formed into a foil is bombarded with thermalneutrons in the presence of deuterium-tritium gas. The resulting fission fragments impart energy to accelerate deuterium and tritium particles which in turn provide approximately 14 MeV neutrons by the reactions t(d,n).sup.4 He and d(t,n).sup.4 He.

An improved method is disclosed for delivering thermalneutrons to a subsurface cancer or tumor which has been first doped with a dopant having a high cross section for neutron capture. The improvement is the use of a guide tube in cooperation with a capillary neutron focusing apparatus, or neutron focusing lens, for directing neutrons to the tumor, and thereby avoiding damage to surrounding tissue. 1 fig.

We have measured the neutron beam characteristics of neutron moderators at the Manuel Lujan Jr. Neutron Scattering Center at LANSCE. The absolute thermalneutron flux, energy spectra and time emission spectra were measured for the high resolution and high intensity decoupled water, partially coupled liquid hydrogen and partially coupled water moderators. The results of our experimental study will provide an insight into aging of different target-moderator-reflector-shield components as well as new experimental data for benchmarking of neutron transport codes.

. Introduction A new type of "bre detector was developed at Nagoya University recently [1}3] and tested of these two materials are rather similar, the quartz "bre is superior regarding the radiation level in which, development of such a detector is of clear importance, and the measurements reported in this paper contribute

We review current state of neutron star cooling theory and discuss the prospects to constrain the equation of state, neutrino emission and superfluid properties of neutron star cores by comparing the cooling theory with observations of thermal radiation from isolated neutron stars.

SHARP Neutronics Expanded SHARP Neutronics Expanded SHARP Neutronics Expanded January 29, 2013 - 1:28pm Addthis Fully heterogeneous predictions of thermalneutron flux in a hypothetical metal-oxide-fueled PWR Fully heterogeneous predictions of thermalneutron flux in a hypothetical metal-oxide-fueled PWR SHARP neutronics Module Development The SHARP neutronics module, PROTEUS, includes neutron and gamma transport solvers and cross-section processing tools as well as the capability for depletion and fuel cycle analysis. The existing high-fidelity solver package was extended to be independent of reactor technology and demonstrated with 2-D MOC and Sn method simulations of LWR core configurations. Efforts to support verification and validation of the DeCART code, used as one reference solution method by the SHARP code

The electron energy spectra, not connected to b-decay, of 235U- and 239Pu-films, irradiated by thermalneutrons, obtained by a Monte Carlo method is presented in the given work. The modelling was performed with the help of a computer code MCNP4C (Monte Carlo Neutron Photon transport code system), allowing to carry out the computer experiments on joint transport of neutrons, photons and electrons. The experiment geometry and the parameters of an irradiation were the same, as in [11] (the thickness of a foil varied only). As a result of computer experiments, the electron spectra was obtained for the samples of 235U, 239Pu and uranium dioxide of 93 % enrichment representing a set of films of 22 mm in diameter and different thickness: 0,001 mm, 0,005 mm, 0,02 mm, 0,01 mm, 0,1 mm, 1,0 mm; and also for the uranium dioxide film of 93 % enrichment (diameter 22 mm and thickness 0,01mm), located between two protective 0,025 mm aluminium disks (the conditions of experiment in [11]) and the electron spectrum was fixed at the output surface of a protective disk. The comparative analysis of the experimental [11] and calculated b--spectra is carried out.

The electron energy spectra, not connected to b-decay, of 235U- and 239Pu-films, irradiated by thermalneutrons, obtained by a Monte Carlo method is presented in the given work. The modelling was performed with the help of a computer code MCNP4C (Monte Carlo Neutron Photon transport code system), allowing to carry out the computer experiments on joint transport of neutrons, photons and electrons. The experiment geometry and the parameters of an irradiation were the same, as in [11] (the thickness of a foil varied only). As a result of computer experiments, the electron spectra was obtained for the samples of 235U, 239Pu and uranium dioxide of 93 % enrichment representing a set of films of 22 mm in diameter and different thickness: 0,001 mm, 0,005 mm, 0,02 mm, 0,01 mm, 0,1 mm, 1,0 mm; and also for the uranium dioxide film of 93 % enrichment (diameter 22 mm and thickness 0,01mm), located between two protective 0,025 mm aluminium disks (the conditions of experiment in [11]) and the electron spectrum was fixed at...

Single-crystal neutron diffraction measures the elastic Bragg reflection intensities from crystals of a material, the structure of which is the subject of investigation. A single crystal is placed in a beam of neutrons produced at a nuclear reactor or at a proton accelerator-based spallation source. Single-crystal diffraction measurements are commonly made at thermalneutron beam energies, which correspond to neutron wavelengths in the neighborhood of 1 Angstrom. For high-resolution studies requiring shorter wavelengths (ca. 0.3-0.8 Angstroms), a pulsed spallation source or a high-temperature moderator (a ''hot source'') at a reactor may be used. When complex structures with large unit-cell repeats are under investigation, as is the case in structural biology, a cryogenic-temperature moderator (a ''cold source'') may be employed to obtain longer neutron wavelengths (ca. 4-10 Angstroms). A single-crystal neutron diffraction analysis will determine the crystal structure of the material, typically including its unit cell and space group, the positions of the atomic nuclei and their mean-square displacements, and relevant site occupancies. Because the neutron possesses a magnetic moment, the magnetic structure of the material can be determined as well, from the magnetic contribution to the Bragg intensities. This latter aspect falls beyond the scope of the present unit; for information on magnetic scattering of neutrons see Unit 14.3. Instruments for single-crystal diffraction (single-crystal diffractometers or SCDs) are generally available at the major neutron scattering center facilities. Beam time on many of these instruments is available through a proposal mechanism. A listing of neutron SCD instruments and their corresponding facility contacts is included in an appendix accompanying this unit.

An invention is described that enables the quantitative simultaneous identification of the matrix materials in which fertile and fissile nuclides are embedded to be made along with the quantitative assay of the fertile and fissile materials. The invention also enables corrections for any absorption of neutrons by the matrix materials and by the measurement apparatus by the measurement of the prompt and delayed neutron flux emerging from a sample after the sample is interrogated by simultaneously applied neutrons and gamma radiation. High energy electrons are directed at a first target to produce gamma radiation. A second target receives the resulting pulsed gamma radiation and produces neutrons from the interaction with the gamma radiation. These neutrons are slowed by a moderator surrounding the sample and bathe the sample uniformly, generating second gamma radiation in the interaction. The gamma radiation is then resolved and quantitatively detected, providing a spectroscopic signature of the constituent elements contained in the matrix and in the materials within the vicinity of the sample. (LEW)

Neutron Basics Neutron Basics A neutron is one of the fundamental particles that make up matter. This uncharged particle exists in the nucleus of a typical atom, along with its positively charged counterpart, the proton. Protons and neutrons each have about the same mass, and both can exist as free particles away from the nucleus. In the universe, neutrons are abundant, making up more than half of all visible matter. Find Out What a Neutron Is Youtube icon Properties of Neutrons How Can Neutrons Be Used for Research? Image of glucose movement in plants Neutron imaging techniques have been able to determine the precise movement of glucose in plants. This knowledge can help scientists better understand how biomass can be efficiently converted into fuel. Neutrons have many properties that make them ideal for certain types of

Two different experiments performed in the 8 MWth MELUSINE experimental power pool reactor aimed at analyzing 1 GWd/t spent fuel pellets doped with several actinides. The goal was to measure the averaged neutron induced capture cross section in two very different neutron spectra (a PWR-like and an under-moderated one). This paper summarizes the combined deterministic APOLLO2-stochastic TRIPOLI4 analysis using the JEFF-3.1.1 European nuclear data library. A very good agreement is observed for most of neutron induced capture cross section of actinides and a clear underestimation for the {sup 241}Am(n,{gamma}) as an accurate validation of its associated isomeric ratio are emphasized. Finally, a possible huge resonant fluctuation (factor of 2.7 regarding to the 1=0 resonance total orbital momenta) is suggested for isomeric ratio. (authors)

Sample records for thermal neutron detector from the National Library of Energy Beta (NLEBeta)

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A thermal radiation shield for cooled portable gamma-ray spectrometers. The thermal radiation shield is located intermediate the vacuum enclosure and detector enclosure, is actively driven, and is useful in reducing the heat load to mechanical cooler and additionally extends the lifetime of the mechanical cooler. The thermal shield is electrically-powered and is particularly useful for portable solid-state gamma-ray detectors or spectrometers that dramatically reduces the cooling power requirements. For example, the operating shield at 260K (40K below room temperature) will decrease the thermal radiation load to the detector by 50%, which makes possible portable battery operation for a mechanically cooled Ge spectrometer.

Experimental techniques are described for the spectral measurement of a collimated fast-neutron beam. A H/sub 2-/ filled cloud chamber, proton-recording nuclear plates, and threshold fission foils were used as neutrondetectors in the measurements. As an application of these techniques, the energy distribution and absolute flux of the fast neutron beam emerging from the Los Alamos fast reactor was measured from 0.1 to 18 Mev. (D.E.B.)

The {sup 3}He neutron spectrometer used for measuring ion temperatures and the NE213 proton recoil spectrometer used for triton burnup measurements were absolutely calibrated with DT and DD neutron generators placed inside the TFTR vacuum vessel. The details of the detector response and calibration are presented. Comparisons are made to the neutron source strengths measured from other calibrated systems. 23 refs., 19 figs., 6 tabs.

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis was performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff versus effective full power days (EFPDs) between the HEU and the LEU cores. The MCNP ATR 1/8th core model was used to optimize the U 235 loading in the LEU core, such that the differences in K-eff and heat flux profiles between the HEU and LEU cores were minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the ATR reference HEU case study. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, the proposed LEU (U-10Mo) core conversion case with nominal fuel meat thickness of 0.330 mm (13 mil) and U-235 enrichment of 19.7 wt% is used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.0 mil) to 0.330 mm (13.0 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). A 0.8g of Boron-10, a burnable absorber, was added in the inner and outer plates to reduce the initial excess reactivity, and the peak to average ratio of the inner/outer heat flux more effectively. Because the B-10 (n,a) reaction will produce Helium-4 (He-4), which might degrade the LEU foil type fuel performance, an alternative absorber option is proposed. The proposed LEU case study will have 6.918 g of Cadmium (Cd) mixed with the LEU at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19) as a burnable absorber to achieve peak to average ratios similar to those for the ATR reference HEU case study.

The Neutron Backscattering technique is tested when performing the task of localizing hydrogenated explosives hidden in soil. Detector system, landmine, soil and neutron source are simulated with Geant4 in order to obtain the number of neutrons detected when several parameters like mine composition, relative position mine-source and soil moisture are varied.0.

Special Nuclear Material (SNM) can either spontaneously fission, or be induced to do so. Either case results in neutron emission. Since neutrons are highly penetrating and difficult to shield, they could, potentially, be detected escaping even a well shielded cargo container. Obviously, if the shielding is sophisticated, detecting it would require a highly efficient detector with close to 4{pi} solid angle coverage. Water Cerenkov detectors may be a cost effective way to achieve that goal if it can be shown that the neutron capture signature is large enough and if sufficient background rejection can be employed as detectors get larger. In 2008 the LLNL Advanced Detector Group reported the successful detection of neutrons with a 1/4 ton gadolinium doped water Cerenkov prototype. We have now built a 4 ton version. This detector is not only bigger, it was designed with photon detection efficiency in mind from the beginning. We are employing increased photocathode coverage and more reflective walls, coated with PTFE. The increased efficiency should allow better energy resolution. We expect that the better diffusive wall reflectivity will reduce the overall dependence of the detector response on particle direction, again producing a more consistent response. We also believe that as detectors get larger, both uncorrelated and correlated backgrounds due to gamma-rays and cosmic ray interactions near the detector will increase. To prove the effectiveness of the technology we must develop new ways to reject these backgrounds while maintaining our sensitivity to SNM neutrons. Better energy resolution will enable us to reject more of the low energy gamma-ray backgrounds on this basis. Overcoming cosmic ray induced neutrons is perhaps an even larger concern as detectors get larger. Our detector is designed so that we can test various segmentation schemes - effectively dividing the detector up into smaller ones. In this presentation, we will describe our detector in detail.

We have designed a portable neutrondetector for passive neutron scanning measurement and coincidence counting of bulk samples of plutonium. The counter will be used for neutron survey applications as well as the measurement of plutonium samples for portable applications. The detector uses advanced design {sup 3}He tubes to increase the efficiency and battery operated shift register electronics. This report describes the hardware, performance, and calibration for the system.

A consistent set of best values of the 2200 meter/second neutron cross sections, Westcott g-factors, and fission neutron yields for /sup 233/U, /sup 235/U, /sup 239/Pu and /sup 241/Pu are presented. A least squares fitting program, LSF, is used to obtain the best fit and to estimate the sensitivity of these fissile parameters to the quoted uncertainties in experimental data. The half-lives of the uranium and plutonium nuclides have been evaluated and these have been used to reassess the significant experimental data. The latest revision of the spontaneous fission neutron yield anti nu, of /sup 252/Cf and the foil thickness corrections to the fission neutron yield ratios of fissile nuclei to /sup 252/Cf are included. These lead to greater consistency in the data used for anti nu (/sup 252/Cf). Similarly, the /sup 234/U half-life as revised leads to improved consistency in the /sup 235/U fission cross section. Comparison is made with the values from ENDF/B-V and other evaluations.

A method of detecting an activator, the method including impinging with an activator a receptor material that includes a photoluminescent material that generates infrared radiation and generating a by-product of a nuclear reaction due to the activator impinging the receptor material. The method further includes generating light from the by-product via the Cherenkov effect, wherein the light activates the photoluminescent material so as to generate the infrared radiation. Identifying a characteristic of the activator based on the infrared radiation.

The Drift Scale Test (DST) is one of the thermal tests being conducted in the Exploratory Studies Facility (ESF). One of the objectives of the DST is to study the coupled thermal-mechanical- hydrological-chemical (TMHC) processes in the ESF at the repository horizon of the potential high-level nuclear waste repository in Yucca Mountain, Nevada. The objectives, the test design, and the test layouts of the DST are included in the test design report by CRWMS M&O Contractor LLNL. The configuration of the DST includes a declining Observation Drift driven mostly east and downward from main tunnel in the ESF, at about 2.827 km from the North portal. The downward slope of the Observation Drift (11.5 to 14.0 percent) ensures a minimum 10 m of middle nonlithophysal Topopah Spring Tuff as the overburden for the DST. The length of the Observation Drift is about 136 m. At the elevation of the DST crown (nominally 10 m below the upper extent of the middle nonlithophysal Topopah Spring Tuff) the Connecting Drift breaks out to the north from the Observation Drift, 136 m from the main tunnel of the ESF. The Connecting Drift extends approximately 40 m to the north from the Observation Drift. A Heater Drift breaks out westward from the Connecting Drift at about 30 m from the Observation Drift. The Heater Drift consists of an 11 m long entry, which includes a plate- loading niche, and a 47 m long heated drift. The nominal diameter of the drifts is 5 m. The detail configuration of the DST, including diagrams showing the drift and borehole layout, is included in the test design report by CRWMS M&O Contractor LLNL. Thermalneutron logging is a method used to determine moisture content in rocks and soils and will be used to monitor moisture content in boreholes ESF-HD-NEU-1 to ESF-HD-NEU-10 (Boreholes 47 to 51 and 64 to 68), ESF-HD-TEMP-1 (Borehole 79), and ESF-HD-TEMP-2 (Borehole 80) in the DST. The neutron probe contains a source of high energy neutrons and a detector for slow (thermal) neutrons. Water present in rocks slows down the neutrons making them detectable (because of the presence of hydrogen).

A well established program of neutron-induced fission cross section measurement at Los Alamos Neutron Science Center (LANSCE) is supporting the Fuel Cycle Research program (FC R&D). The incident neutron energy range spans from sub-thermal up to 200 MeV by combining two LANSCE facilities, the Lujan Center and the Weapons Neutron Research center (WNR). The time-of-flight method is implemented to measure the incident neutron energy. A parallel-plate fission ionization chamber was used as a fission fragment detector. The event rate ratio between the investigated foil and a standard {sup 235}U foil is translated into a fission cross section ratio. Thin actinide targets with deposits of <200 {micro}g/cm{sup 2} on stainless steel backing were loaded into a fission chamber. In addition to previously measured data for {sup 237}Np, {sup 239-242}Pu, {sup 243}Am, new measurements include the recently completed {sup 233,238}U isotopes, {sup 236}U data which is being analyzed, and {sup 234}U data acquired in the 2011-2012 LANSCE run cycle. The new data complete the full suite of Uranium isotopes which were investigated with this experimental approach. When analysis of the new measured data is completed, data will be delivered to evaluators. Having data for multiple Uranium isotopes will support theoretical modeling capabilities and strengthens nuclear data evaluation.

A segmented neutron calorimeter using nine 4-inch x 4-inch x 48-inch plastic scintillators and sixteen 2-inch-diameter 48-inch-long 200-mbar-{sup 3}He drift tubes is described. The correlated scintillator and neutron-capture events provide a means for n/{gamma} discrimination, critical to the neutron calorimetry when the {gamma} background is substantial and the {gamma} signals are comparable in amplitude to the neutron signals. A single-cell prototype was constructed and tested. It can distinguish between a {sup 17}N source and a {sup 252}Cf source when the {gamma} and the thermalneutron background are sufficiently small. The design and construction of the nine-cell segmented detector assembly follow the same principle. By recording the signals from individual scintillators, additional {gamma}-subtraction schemes, such as through the time-of-flight between two scintillators, may also be used. The variations of the light outputs from different parts of a scintillator bar are less than 10%.

Cosmic rays interact with the surface of a planetary body and produce a cascade of secondary particles, such as neutrons. Neutron-induced scattering and capture reactions play an important role in the production of discrete gamma-ray lines that can be measured by a gamma-ray spectrometer on board of an orbiting spacecraft. These data can be used to determine the concentration of many elements in the surface of a planetary body, which provides clues to its bulk composition and in turn to its origin and evolution. To investigate the gamma rays made by neutron interactions, thin targets were irradiated with neutrons having energies from 14 MeV to 0.025 eV. By means of foil activation technique the ratio of epithermal to thermalneutrons was determined to be similar to that in the Moon. Gamma rays emitted by the targets and the surrounding material were detected by a high-resolution germanium detector in the energy range of 0.1 to 8 MeV. Most of the gamma-ray lines that are expected to be used for planetary gamma-ray spectroscopy were found in the recorded spectra and the principal lines in these spectra are presented. 58 refs., 7 figs., 9 tabs.

A method of measuring neutron radiation within a nuclear reactor is provided. A sintered oxide wire is disposed within the reactor and exposed to neutron radiation. The induced radioactivity is measured to provide an indication of the neutron energy and flux within the reactor.

We propose modifying large water \\v{C}erenkov detectors by the addition of 0.2% gadolinium trichloride, which is highly soluble, newly inexpensive, and transparent in solution. Since Gd has an enormous cross section for radiative neutron capture, with $\\sum E_\\gamma = 8$ MeV, this would make neutrons visible for the first time in such detectors, allowing antineutrino tagging by the coincidence detection reaction $\\bar{\

A phase detector circuit is described for use at very high frequencies of the order of 50 megacycles. The detector circuit includes a pair of rectifiers inverted relative to each other. One voltage to be compared is applied to the two rectifiers in phase opposition and the other voltage to be compared is commonly applied to the two rectifiers. The two result:ng d-c voltages derived from the rectifiers are combined in phase opposition to produce a single d-c voltage having amplitude and polarity characteristics dependent upon the phase relation between the signals to be compared. Principal novelty resides in the employment of a half-wave transmission line to derive the phase opposing signals from the first voltage to be compared for application to the two rectifiers in place of the transformer commonly utilized for such purpose in phase detector circuits for operation at lower frequency.

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A microwave detector is provided for measuring the envelope shape of a microwave pulse comprised of high-frequency oscillations. A biased ferrite produces a magnetization field flux that links a B-dot loop. The magnetic field of the microwave pulse participates in the formation of the magnetization field flux. High-frequency insensitive means are provided for measuring electric voltage or current induced in the B-dot loop. The recorded output of the detector is proportional to the time derivative of the square of the envelope shape of the microwave pulse.

A microwave detector (10) is provided for measuring the envelope shape of a microwave pulse comprised of high-frequency oscillations. A biased ferrite (26, 28) produces a magnetization field flux that links a B-dot loop (16, 20). The magnetic field of the microwave pulse participates in the formation of the magnetization field flux. High-frequency insensitive means (18, 22) are provided for measuring electric voltage or current induced in the B-dot loop. The recorded output of the detector is proportional to the time derivative of the square of the envelope shape of the microwave pulse.

7 7 Staff Awards: 2007 Chakoumakos elected MSA Fellow Bryan Chakoumakos Neutron scientist Bryan Chakoumakos was recently elected a fellow of the Mineralogical Society of America. A member of the Neutron Scattering Science Division, Bryan leads the Single-Crystal Diffraction Group. The group has five neutron scattering instruments in various stages of design and construction, located at HFIR and SNS. The MSA was founded in 1919 and, among other goals, encourages fundamental research on natural materials and supports education through its publications, educational grants, and courses. Pharos NeutronDetector System Researchers at the Department of Energy's Oak Ridge National Laboratory have won six R&D 100 Awards, given annually by R&D Magazine to the year's

A modular design is proposed for an electron antineutrino detector based on boron-doped liquid scintillator. Tests have been carried out on small detector systems using neutrons to simulate the antineutrino detection signature. Results from these tests are reported, and the possibility of using a larger system of similar design to detect reactor antineutrinos is discussed.

The present invention is for a radiation detector apparatus for detecting radiation sources present in cargo shipments. The invention includes the features of integrating a bubble detector sensitive to neutrons and a GPS system into a miniaturized package that can wirelessly signal the presence of radioactive material in shipping containers. The bubble density would be read out if such indicated a harmful source.

An experiment to measure the time reversal invariance violating (T-violating) triple correlation (D) in the decay of free polarized neutrons has been developed. The detector design incorporates a detector geometry that provides a significant improvement in the sensitivity over that used in the most sensitive of previous experiments. A prototype detector was tested in measurements with a cold neutron beam. Data resulting from the tests are presented. A detailed calculation of systematic effects has been performed and new diagnostic techniques that allow these effects to be measured have been developed. As the result of this work, a new experiment is under way that will improve the sensitivity to D to 3 {times} 10{sup {minus}4} or better. With higher neutron flux a statistical sensitivity of the order 3 {times} 10{sup {minus}5} is ultimately expected. The decay of free polarized neutrons (n {yields} p + e + {bar v}{sub e}) is used to search for T-violation by measuring the triple correlation of the neutron spin polarization, and the electron and proton momenta ({sigma}{sub n} {center_dot} p{sub p} {times} p{sub e}). This correlation changes sign under reversal of the motion. Since final state effects in neutron decay are small, a nonzero coefficient, D, of this correlation indicates the violation of time reversal invariance. D is measured by comparing the numbers of coincidences in electron and proton detectors arranged symmetrically about a longitudinally polarized neutron beam. Particular care must be taken to eliminate residual asymmetries in the detectors or beam as these can lead to significant false effects. The Standard Model predicts negligible T-violating effects in neutron decay. Extensions to the Standard Model include new interactions some of which include CP-violating components. Some of these make first order contributions to D.

The countrate response of a gamma spectrometry system from a neutron radiation source behind a plane of moderating material doped with a nuclide of a large radiative neutron capture cross-section exhibits a countrate response analogous to a gamma radiation source at the same position from the detector. Using a planar, surface area of the neutron moderating material exposed to the neutron radiation produces a larger area under the prompt gamma ray peak in the detector than a smaller area of dimensions relative to the active volume of the gamma detection system.

We studied Bragg diffraction and Thermal Diffuse Scattering (TDS) from a Si(111) channel-cut triple-bounce crystal using the time-of-flight technique at a pulsed neutron source. Cadmium shielding restricted the detector s direct view of the first bounce surface. The channel-cut crystal dramatically suppresses TDS in the vicinity of the (111), (333) and (444) Bragg reflections; however, TDS appears and increases with the decrease of wavelength in the range of the (555), (777) and (888) orders where cadmium becomes transparent and the single-bounce reflections and TDS contaminate the triple-bounce (555), (777) and (888) reflections.

Radiation portal monitors used for interdiction of illicit materials at borders include highly sensitive neutron detection systems. The main reason for having neutron detection capability is to detect fission neutrons from plutonium. The currently deployed radiation portal monitors (RPMs) from Ludlum and Science Applications International Corporation (SAIC) use neutrondetectors based upon 3He-filled gas proportional counters, which are the most common large neutrondetector. There is a declining supply of 3He in the world, and thus, methods to reduce the use of this gas in RPMs with minimal changes to the current system designs and sensitivity to cargo-borne neutrons are being investigated. Four technologies have been identified as being currently commercially available, potential alternative neutrondetectors to replace the use of 3He in RPMs. These technologies are: 1) Boron trifluoride (BF3)-filled proportional counters, 2) Boron-lined proportional counters, 3) Lithium-loaded glass fibers, and 4) Coated non-scintillating plastic fibers. In addition, a few other companies have detector technologies that might be competitive in the near term as an alternative technology. Reported here are the results of tests of a boron-lined, multitube proportional counter manufactured by Centronic Ltd. (Surry, U.K. and Houston, TX). This testing measured the required performance for neutron detection efficiency and gamma-ray rejection capabilities of the detector.

Scattering Scattering Neutron Scattering Facilities at HFIR The fully instrumented HFIR will eventually include 15 state-of-the-art neutron scattering instruments, seven of which will be designed exclusively for cold neutron experiments, located in a guide hall south of the reactor building. The currently available instruments and the status of new instruments can be found on the HFIR Instrument Systems pages. Particularly prominent in the cold neutron guide hall are the two small-angle neutron scattering (SANS) instruments, each terminating in a 70-ft-long evacuated cylinder containing a large moveable neutrondetector. In addition to the instruments, laboratories are equipped for users to prepare samples. Perhaps the most exciting development at HFIR is the successfully

A research program has been initiated by the NEET program for developing and testing compact miniature fission chambers capable of simultaneously measuring thermalneutron flux, fast neutron flux and temperature within a single package. When implemented, these sensors will significantly advance flux detection capabilities for irradiation tests in US Materials Test Reactors (MTRs).Ultimately, evaluations may lead to a more compact, more accurate, and longer lifetime flux sensor for critical mock-ups, high performance reactors and commercial nuclear power plants. Deployment of Micro-Pocket Fission Detectors (MPFDs) in US DOE-NE program irradiation tests will address several challenges: Current fission chamber technologies do not offer the ability to measure fast flux, thermal flux and temperature within a single compact probe, MPFDs offer this option. MPFD construction is very different then current fission chamber construction; the use of high temperature materials allow MPFDs to be specifically tailored to survive harsh conditions in typical high performance MTR irradiation tests. New high-fidelity reactor physics codes will need a small, accurate, multipurpose in-core sensor to validate the codes without perturbing the validation experiment; MPFDs fill this requirement. MPFDs can be built with variable sensitivities to survive the lifetime of an experiment or fuel assembly in some MTRs; allowing for more efficient and cost effective power monitoring. The small size of the MPFDs allows multiple sensors to be simultaneously deployed; obtaining data required to visualize the reactor flux and temperature profiles. This report summarizes the research progress for year 1 of this 3 year project. An updated design of the MPFD has been developed, materials and tools to support the new design have been procured, construction methods to support the new design have been initiated at INLs HTTL and KSUs SMART Laboratory, plating methods are being updated at KSU, new detector electronics have been designed, built and tested at KSU. In addition, a project meeting was held at KSU and a detector evaluation plan has been initiated between INL and KSU. Once NEET program evaluations are completed, the final MPFD will be deployed in MTR irradiations, enabling DOE-NE programs evaluating the performance of candidate new fuels and materials to better characterize irradiation test conditions.

Capabilities of the ARCS Instrument Capabilities of the ARCS Instrument ARCS Overview The wide angular-range chopper spectrometer ARCS at the Spallation Neutron Source (SNS) is optimized to provide a high neutron flux at the sample position with a large solid angle of detector coverage. The instrument incorporates modern neutron instrumentation, such as an elliptically focused neutron guide, high speed magnetic bearing choppers, and a massive array of 3He linear position sensitive detectors. Novel features of the spectrometer include the use of a large gate valve between the sample and detector vacuum chambers and the placement of the detectors within the vacuum, both of which provide a window-free final flight path to minimize background scattering while allowing rapid changing of the sample and

A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

A neutron source which is particularly useful for neutron radiography consists of a vessel containing a moderating media of relatively low moderating ratio, a flux trap including a moderating media of relatively high moderating ratio at the center of the vessel, a shell of depleted uranium dioxide surrounding the moderating media of relatively high moderating ratio, a plurality of guide tubes each containing a movable source of neutrons surrounding the flux trap, a neutron shield surrounding one part of each guide tube, and at least one collimator extending from the flux trap to the exterior of the neutron source. The shell of depleted uranium dioxide has a window provided with depleted uranium dioxide shutters for each collimator. Reflectors are provided above and below the flux trap and on the guide tubes away from the flux trap.

Neutrondetectors are commonly used by the nuclear materials processing industry to monitor fissile materials in process vessels and tanks. The proper functioning of these neutron monitors must be periodically evaluated. We have developed and placed in routine use a PC-based multichannel analyzer (MCA) system for on-line BF3 and He-3 gas-filled detector function testing. The automated system: 1) acquires spectral data from the monitor system, 2) analyzes the spectrum to determine the detector`s functionality, 3) makes suggestions for maintenance or repair, as required, and 4) saves the spectrum and results to disk for review. The operator interface has been designed to be user-friendly and to minimize the training requirements of the user. The system may also be easily customized for various applications

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NEUTRON CAPTURE LOGS IN THINLY-BEDDED FORMATIONS Jordan G. Mimoun and Carlos Torres-VerdĂ­n, The University to capture neutrons. The lower the neutron energy, the more likely capture phenomena will take place; hence neutrons at thermal energies are the most likely to be absorbed. Consequently, monitoring the population

The neutron long counter NERO was built at the National Superconducting Cyclotron Laboratory (NSCL), Michigan State University, for measuring beta-delayed neutron-emission probabilities. The detector was designed to work in conjunction with a beta-decay implantation station, so that beta decays and beta-delayed neutrons emitted from implanted nuclei can be measured simultaneously. The high efficiency of about 40%, for the range of energies of interest, along with the small background, are crucial for measuring beta-delayed neutron emission branchings for neutron-rich r-process nuclei produced as low intensity fragmentation beams in in-flight separator facilities.

We report on the neutron response of the LAMBDA spectrometer developed earlier for high-energy gamma-ray measurement. The energy dependent neutron detection efficiency of the spectrometer has been measured using the time-of-flight (TOF) technique and compared with that of an organic liquid scintillator based neutrondetector (BC501A). The extracted efficiencies have also been compared with those obtained from Monte Carlo GEANT4 simulation. We have also measured the average interaction length of neutrons in the BaF2 crystal in a separate experiment, in order to determine the TOF energy resolution. Finally, the LAMBDA spectrometer has been tested in an in-beam-experiment by measuring neutron energy spectra in the 4He + 93Nb reaction to extract nuclear level density parameters. Nuclear level density parameters obtained by the LAMBDA spectrometer were found to be consistent with those obtained by the BC501A neutrondetector, indicating that the spectrometer can be efficiently used as a neutrondetector to measure the nuclear level density parameter.

During the past several years, there has been growing interest in Boron Neutron Capture Therapy (BNCT) using epithermal neutron beams. The dosimetry of these beams is challenging. The incident beam is comprised mostly of epithermal neutrons, but there is some contamination from photons and fast neutrons. Within the patient, the neutron spectrum changes rapidly as the incident epithermal neutrons scatter and thermalize, and a photon field is generated from neutron capture in hydrogen. In this paper, a method to determine the doses from thermal and fast neutrons, photons, and the B-10([ital n],[alpha])Li-7 reaction is presented. The photon and fast neutron doses are measured with ionization chambers, in realistic phantoms, using the dual chamber technique. The thermalneutron flux is measured with gold foils using the cadmium difference technique; the thermalneutron and B-10 doses are determined by the kerma factor method. Representative results are presented for a unilateral irradiation of the head. Sources of error in the method as applied to BNCT dosimetry, and the uncertainties in the calculated doses are discussed.

University Of Richmond - Department of Physics Software We simulated the neutron detection efficiency physics program. Specifically, we are simulating the neutron detection efficiency of the forward TOFSimulating the Neutron Detection Efficiency of the CLAS12 Detector M. Moog and G. Gilfoyle

A neutron range spectrometer and method for determining the neutron energy spectrum of a neutron emitting source are disclosed. Neutrons from the source are colliminated along a collimation axis and a position sensitive neutron counter is disposed in the path of the collimated neutron beam. The counter determines positions along the collimation axis of interactions between the neutrons in the neutron beam and a neutron-absorbing material in the counter. From the interaction positions, a computer analyzes the data and determines the neutron energy spectrum of the neutron beam. The counter is preferably shielded and a suitable neutron-absorbing material is He-3. 1 fig.

A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

A neutron source is obtained without employing any separate beryllia receptacle, as was formerly required. The new method is safer and faster, and affords a source with both improved yield and symmetry of neutron emission. A Be container is used to hold and react with Pu. This container has a thin isolating layer that does not obstruct the desired Pu--Be reaction and obviates procedures previously employed to disassemble and remove a beryllia receptacle. (AEC)

(cont.) A storage system was designed to contain the lithium-6 filter safely when it is not in use. A mixed field dosimetry method was used to measure the photon, thermalneutron and fast neutron dose. The measured advantage ...

It is now possible to model thermal relaxation of neutron stars after bouts of accretion during which the star is heated out of equilibrium by nuclear reactions in its crust. Major uncertainties in these models can be encapsulated in modest variations of a handful of fudge parameters that change the crustal thermal conductivity, specific heat, and heating rates. Observations of thermal relaxation constrain these fudge parameters and allow us to predict longer term variability in terms of the neutron star core temperature. We demonstrate this explicitly by modeling ongoing thermal relaxation in the neutron star XTE J1701-462. Its future cooling, over the next 5 to 30 years, is strongly constrained and depends mostly on its core temperature, uncertainties in crust physics having essentially been pinned down by fitting to the first three years of observations.

A neutron imaging diagnostic has recently been commissioned at the National Ignition Facility (NIF). This new system is an important diagnostic tool for inertial fusion studies at the NIF for measuring the size and shape of the burning DT plasma during the ignition stage of ICF implosions. The imaging technique utilizes a pinhole neutron aperture, placed between the neutron source and a neutrondetector. The detection system measures the two dimensional distribution of neutrons passing through the pinhole. This diagnostic has been designed to collect two images at two times. The long flight path for this diagnostic, 28 m, results in a chromatic separation of the neutrons, allowing the independently timed images to measure the source distribution for two neutron energies. Typically the first image measures the distribution of the 14 MeV neutrons and the second image of the 6-12 MeV neutrons. The combination of these two images has provided data on the size and shape of the burning plasma within the compressed capsule, as well as a measure of the quantity and spatial distribution of the cold fuel surrounding this core.

A neutron imaging diagnostic has recently been commissioned at the National Ignition Facility (NIF). This new system is an important diagnostic tool for inertial fusion studies at the NIF for measuring the size and shape of the burning DT plasma during the ignition stage of Inertial Confinement Fusion (ICF) implosions. The imaging technique utilizes a pinhole neutron aperture, placed between the neutron source and a neutrondetector. The detection system measures the two dimensional distribution of neutrons passing through the pinhole. This diagnostic has been designed to collect two images at two times. The long flight path for this diagnostic, 28 m, results in a chromatic separation of the neutrons, allowing the independently timed images to measure the source distribution for two neutron energies. Typically the first image measures the distribution of the 14 MeV neutrons and the second image of the 6-12 MeV neutrons. The combination of these two images has provided data on the size and shape of the burning plasma within the compressed capsule, as well as a measure of the quantity and spatial distribution of the cold fuel surrounding this core.

The MEGAPIE project is one of the key experiments towards the feasibility of Accelerator Driven Systems. On-line operation and post-irradiation analysis will provide the scientific community with unique data on the behavior of a liquid spallation target under realistic irradiation conditions. A good neutronics performance of such a target is of primary importance towards an intense neutron source, where an extended liquid metal loop requires some dedicated verifications related to the delayed neutron activity of the irradiated PbBi. In this paper we report on the experimental characterization of the MEGAPIE neutronics in terms of the prompt neutron (PN) flux inside the target and the delayed neutron (DN) flux on the top of it. For the PN measurements, a complex detector, made of 8 microscopic fission chambers, has been built and installed in the central part of the target to measure the absolute neutron flux and its spatial distribution. Moreover, integral information on the neutron energy distribution as a function of the position along the beam axis could be extracted, providing integral constraints on the neutron production models implemented in transport codes such as MCNPX. For the DN measurement, we used a standard 3He counter and we acquired data during the start-up phase of the target irradiation in order to take sufficient statistics at variable beam power. Experimental results obtained on the PN flux characteristics and their comparison with MCNPX simulations are presented, together with a preliminary analysis of the DN decay time spectrum.

We present the results of an Ultracold neutron (UCN) production experiment in a pulsed neutron beam line at the Los Alamos Neutron Scattering Center. The experimental apparatus allows for a comprehensive set of measurements of UCN production as a function of target temperature, incident neutron energy, target volume, and applied magnetic field. However, the low counting statistics of the UCN signal expected can be overwhelmed by the large background associated with the scattering of the primary cold neutron flux that is required for UCN production. We have developed a background subtraction technique that takes advantage of the very different time-of-flight profiles between the UCN and the cold neutrons, in the pulsed beam. Using the unique timing structure, we can reliably extract the UCN signal. Solid ortho-D$_2$ is used to calibrate UCN transmission through the apparatus, which is designed primarily for studies of UCN production in solid O$_2$. In addition to setting the overall detection efficiency in the apparatus, UCN production data using solid D$_2$ suggest that the UCN upscattering cross-section is smaller than previous estimates, indicating the deficiency of the incoherent approximation widely used to estimate inelastic cross-sections in the thermal and cold regimes.

Since long time, the compelling scientific goals of future high energy physics experiments were a driving factor in the development of advanced detector technologies. A true innovation in detector instrumentation concepts came in 1968, with the development of a fully parallel readout for a large array of sensing elements - the Multiwire Proportional Chamber (MWPC), which earned Georges Charpak a Nobel prize in physics in 1992. Since that time radiation detection and imaging with fast gaseous detectors, capable of economically covering large detection volume with low mass budget, have been playing an important role in many fields of physics. Advances in photo-lithography and micro-processing techniques in the chip industry during the past decade triggered a major transition in the field of gas detectors from wire structures to Micro-Pattern Gas Detector (MPGD) concepts, revolutionizing cell size limitations for many gas detector applications. The high radiation resistance and excellent spatial and time resolution make them an invaluable tool to confront future detector challenges at the next generation of colliders. The design of the new micro-pattern devices appears suitable for industrial production. Novel structures where MPGDs are directly coupled to the CMOS pixel readout represent an exciting field allowing timing and charge measurements as well as precise spatial information in 3D. Originally developed for the high energy physics, MPGD applications has expanded to nuclear physics, UV and visible photon detection, astroparticle and neutrino physics, neutron detection and medical physics.

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Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the 'user effect' and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users [1]. Five seminars have been held at University of Pisa (2003, 2004), at The Pennsylvania State University (2004), at University of Zagreb (2005) and at the School of Industrial Engineering of Barcelona (2006). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed 'permanent' stepwise approach to user training. The 3D S.UN.COP 2006 was successfully held with the attendance of 33 participants coming from 18 countries and 28 different institutions (universities, vendors, national laboratories and regulatory bodies). More than 30 scientists (coming from 13 countries and 23 different institutions) were involved in the organization of the seminar, presenting theoretical aspects of the proposed methodologies and holding the training and the final examination. A certificate (LA Code User grade) was released to participants that successfully solved the assigned problems. A sixth seminar will be organized in 2007 at the Texas A and M University involving more than 30 scientists between lecturers and code developers. (http://dimnp.ing.unipi.it/3dsuncop/2007). (authors)

Earth is connected gravitationally, magnetically and electrically to its heat source - a neutron star that is obscured from view by waste products in the photosphere. Neutron repulsion is like the hot filament in an incandescent light bulb. Excited neutrons are emitted from the solar core and decay into hydrogen that glows in the photosphere like a frosted light bulb. Neutron repulsion was recognized in nuclear rest mass data in 2000 as the overlooked source of energy, the keystone of an arch that locked together these puzzling space-age observations: 1.) Excess 136Xe accompanied primordial helium in the stellar debris that formed the solar system (Fig. 1); 2.) The Sun formed on the supernova core (Fig. 2); 3.) Waste products from the core pass through an iron-rich mantle, selectively carrying lighter elements and lighter isotopes of each element into the photosphere (Figs. 3-4); and 4.) Neutron repulsion powers the Sun and sustains life (Figs. 5-7). Together these findings offer a framework for understanding how: a.) The Sun generates and releases neutrinos, energy and solar-wind hydrogen and helium; b.) An inhabitable planet formed and life evolved around an ordinary-looking star; c.) Continuous climate change - induced by cyclic changes in gravitational interactions of the Sun's energetic core with planets - has favored survival by adaptation.

The Savannah River Laboratory (SRL), in conjunction with Savannah River Site (SRS) Separations Technology personnel, has developed and implemented a comprehensive program to improve the performance and reliability of neutrondetector systems (neutron monitors) in the SRS separations areas. The neutron monitors, which monitor the buildup of fissile material in the mixer-settler banks of the solvent extraction process, are important nuclear safety control devices. A review of the performance history of the neutron monitors reveals that many of the systems exhibit problems arising from several causes, including: low neutron sensitivity, high susceptibility to electromagnetic interferences (due to long cable runs between detectors and their electronics), and high maintenance requirements. To address these problems, the neutron monitor improvement program encompasses both technical and administrative improvements, including: substitution of more sensitive neutron monitors at many locations in the solvent extraction areas, the development of an integrated preamplifier/amplifier package to eliminate long cable runs, and improvements in the neutron monitor functional test procedures to reduce maintenance requirements. The implementation of these improvements, already partially complete, is expected to provide enhanced operation and reliability for the neutron monitors. This paper will present a description of the solvent neutron monitors as well as technical details of the improvement program. 2 refs., 5 figs., 1 tab.

The Cryogenic Apparatus for Precision Tests of Argon Interactions with Neutrino (CAP- TAIN) program is designed to make measurements of scientific importance to long-baseline neutrino physics and physics topics that will be explored by large underground detectors. The CAPTAIN detector is a liquid argon TPC deployed in a portable and evacuable cryostat. Five tons of liquid argon are instrumented with a 2,000 channel liquid argon TPC and a photon detection system. Subsequent to the commissioning phase, the detector will collect data in a high-energy neutron beamline that is part of the Los Alamos Neutron Science Center to measure cross-sections of spallation products that are backgrounds to measurements of neutrinos from a supernova burst, cross-sections of events that mimic the electron neutrino appearance signal in long-baseline neutrino physics and neutron signatures to constrain neutrino energy reconstruction in LBNE's long-baseline program. Subsequent to the neutron running, the CAPTAIN detector will be moved to a neutrino source. Two possibilities are an on-axis run in the NuMI beamline at FNAL and a run in the neutrino source produced by the SNS. An on-axis run at NuMI produces more than one million events of interest in a two or three year run at neutrino energies between 1 and 10 GeV - complementary to the MicroBooNE experiment, which will measure similar interactions at a lower energy range - 0.5 to 2 GeV. At the SNS the neutrinos result from the decays stopped positively charged pions and muons yielding a broad spectrum up to 50 MeV. If located close to the spallation target, CAPTAIN can detect several thousand events per year in the same neutrino energy regime where neutrinos from a supernova burst are. Measurements at the SNS yield a first measurement of the cross- section of neutrinos on argon in this important energy regime.

A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

A wide variety of neutrondetector systems have been used at various research facilities to search for anomalous neutron emission from deuterated metals. Some of these detector systems are summarized here together with possible sources of spurious signals from electronic noise. During the past two years, we have performed experiments to measure neutron emission from pressurized D{sub 2} gas mixed with various forms of titanium metal chips and sponge. Details concerning the neutrondetectors, experimental procedures, and results have been reported previously. Our recent experiments have focused on increasing the low-level neutron emission and finding a way to trigger the emission. To improve our detection sensitivity, we have increased the shielding in our counting laboratory, changed to low-background {sup 3}He tubes, and set up additional detector systems in deep underground counting stations. This report is an update on this experimental work. 7 refs., 5 figs., 4 tabs.

A device for detecting neutrons includes a semi-insulated bulk semiconductor substrate having opposed polished surfaces. A blocking Schottky contact comprised of a series of metals such as Ti, Pt, Au, Ge, Pd, and Ni is formed on a first polished surface of the semiconductor substrate, while a low resistivity ("ohmic") contact comprised of metals such as Au, Ge, and Ni is formed on a second, opposed polished surface of the substrate. In one embodiment, n-type low resistivity pinout contacts comprised of an Au/Ge based eutectic alloy or multi-layered Pd/Ge/Ti/Au are also formed on the opposed polished surfaces and in contact with the Schottky and ohmic contacts. Disposed on the Schottky contact is a neutron reactive film, or coating, for detecting neutrons. The coating is comprised of a hydrogen rich polymer, such as a polyolefin or paraffin; lithium or lithium fluoride; or a heavy metal fissionable material. By varying the coating thickness and electrical settings, neutrons at specific energies can be detected. The coated neutrondetector is capable of performing real-time neutron radiography in high gamma fields, digital fast neutron radiography, fissile material identification, and basic neutron detection particularly in high radiation fields.

Spontaneous fission in Special Nuclear Material (SNM) such as plutonium and highly enriched uranium (HEU) results in the emission of neutrons with energies in the MeV range (hereafter 'fast neutrons'). These fast neutrons are largely unaffected by the few centimeters of intervening high-Z material that would suffice for attenuating most emitted gamma rays, while tens of centimeters of hydrogenous materials are required to achieve substantial attenuation of neutron fluxes from SNM. Neutrondetectors are therefore an important complement to gamma-ray detectors in SNM search and monitoring applications. The rate at which SNM emits fast neutrons varies from about 2 per kilogram per second for typical HEU to some 60,000 per kilogram per second for metallic weapons grade plutonium. These rates can be compared with typical sea-level (cosmogenic) neutron backgrounds of roughly 5 per second per square meter per steradian in the relevant energy range [1]. The fact that the backgrounds are largely isotropic makes directional neutron detection especially attractive for SNM detection. The ability to detect, localize, and ultimately identify fast neutron sources at standoff will ultimately be limited by this background rate. Fast neutrons are particularly well suited to standoff detection and localization of SNM or other fast neutrons sources. Fast neutrons have attenuation lengths of about 60 meters in air, and retain considerable information about their source direction even after one or two scatters. Knowledge of the incoming direction of a fast neutron, from SNM or otherwise, has the potential to significantly improve signal to background in a variety of applications, since the background arriving from any one direction is a small fraction of the total background. Imaging or directional information therefore allows for source detection at a larger standoff distance or with shorter dwell times compared to nondirectional detectors, provided high detection efficiency can be maintained. Directional detection of neutrons has been previously considered for applications such as controlled fusion neutron imaging [2], nuclear fuel safety research [3], imaging of solar neutrons and SNM [4], and in nuclear science [5]. The use of scintillating crystals and fibers has been proposed for directional neutron detection [6]. Recently, a neutron scatter camera has been designed, constructed, and tested for imaging of fast neutrons, characteristic for SNM material fission [7]. The neutron scatter camera relies on the measurement of the proton recoil angle and proton energy by time of flight between two segmented solid-state detectors. A single-measurement result from the neutron scatter camera is a ring containing the possible incident neutron direction. Here we describe the development and commissioning of a directional neutron detection system based on a time projection chamber (TPC) detector. The TPC, which has been widely used in particle and nuclear physics research for several decades, provides a convenient means of measuring the full 3D trajectory, specific ionization (i.e particle type) and energy of charged particles. For this application, we observe recoil protons produced by fast neutron scatters on protons in hydrogen or methane gas. Gas pressures of a few ATM provide reasonable neutron interaction/scattering rates.

Among the most promising sources of gravitational waves for ground?based detectors are the signals emitted during the coalescence of compact binary systems containing neutron stars or black holes. In recent years

The Spallation Neutron Source (SNS) at Oak Ridge National Laboratory currently holds the Guinness World Record as the world most powerful pulsed spallation neutron source. Neutrons scattered off atomic nuclei in a sample yield important information about the position, motions, and magnetic properties of atoms in materials. A neutron scattering experiment usually involves sample environment control (temperature, pressure, etc.), mechanical alignment (slits, sample and detector position), magnetic field controllers, neutron velocity selection (choppers) and neutrondetectors. The SNS Data Acquisition System (DAS) consists of real-time sub-system (detector read-out with custom electronics, chopper interface), data preprocessing (soft real-time) and a cluster of control and ancillary PCs. The real-time system runs FPGA firmware and programs running on PCs (C++, LabView) typically perform one task such as motor control and communicate via TCP/IP networks. PyDas is a set of Python modules that are used to integrate various components of the SNS DAS system. It enables customized automation of neutron scattering experiments in a rapid and flexible manner. It provides wxPython GUIs for routine experiments as well as IPython command line scripting. Matplotlib and numpy are used for data presentation and simple analysis. We will present an overview of SNS Data Acquisition System and PyDas architectures and implementation along with the examples of use. We will also discuss plans for future development as well as the challenges that have to be met while maintaining PyDas for 20+ different scientific instruments.

The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

A new method for the determination of the multiplication of highly enriched uranium systems is presented. The method uses delayed neutrons to drive the HEU system. These delayed neutrons are from fission events induced by a pulsed 14-MeV neutron source. Between pulses, neutrons are detected within a medium efficiency neutrondetector using {sup 3}He ionization tubes within polyethylene enclosures. The neutron detection times are recorded relative to the initiation of the 14-MeV neutron pulse, and subsequently analyzed with the Feynman reduced variance method to extract singles, doubles and triples neutron counting rates. Measurements have been made on a set of nested hollow spheres of 93% enriched uranium, with mass values from 3.86 kg to 21.48 kg. The singles, doubles and triples counting rates for each uranium system are compared to calculations from point kinetics models of neutron multiplicity to assign multiplication values. These multiplication values are compared to those from MC NP K-Code calculations.

A micro-machined thermal conductivity detector for a portable gas chromatograph. The detector is highly sensitive and has fast response time to enable detection of the small size gas samples in a portable gas chromatograph which are in the order of nanoliters. The high sensitivity and fast response time are achieved through micro-machined devices composed of a nickel wire, for example, on a silicon nitride window formed in a silicon member and about a millimeter square in size. In addition to operating as a thermal conductivity detector, the silicon nitride window with a micro-machined wire therein of the device can be utilized for a fast response heater for PCR applications.

A neutronic reactor comprising an active portion containing material fissionable by neutrons of thermal energy, means to control a neutronic chain reaction within the reactor comprising a safety device and a regulating device, a safety device including means defining a vertical channel extending into the reactor from an aperture in the upper surface of the reactor, a rod containing neutron-absorbing materials slidably disposed within the channel, means for maintaining the safety rod in a withdrawn position relative to the active portion of the reactor including means for releasing said rod on actuation thereof, a hopper mounted above the active portion of the reactor having a door disposed at the bottom of the hopper opening into the vertical channel, a plurality of bodies of neutron-absorbing materials disposed within the hopper, and means responsive to the failure of the safety rod on actuation thereof to enter the active portion of the reactor for opening the door in the hopper.

An improved neutron reflecting supermirror structure comprising a plurality of stacked sets of bilayers of neutron reflecting materials. The improved neutron reflecting supermirror structure is adapted to provide extremely good performance at high incidence angles, i.e. up to four time the critical angle of standard neutron mirror structures. The reflection of neutrons striking the supermirror structure at a high critical angle provides enhanced neutron throughput, and hence more efficient and economical use of neutron sources.

An improved neutron reflecting supermirror structure comprising a plurality of stacked sets of bilayers of neutron reflecting materials. The improved neutron reflecting supermirror structure is adapted to provide extremely good performance at high incidence angles, i.e. up to four time the critical angle of standard neutron mirror structures. The reflection of neutrons striking the supermirror structure at a high critical angle provides enhanced neutron throughput, and hence more efficient and economical use of neutron sources. 2 figs.

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This handbook reviews those problems and methods of science and technology where the neutrons produced in the /sup 3/H/d, n//sup 4/He and /sup 2/H/d, N//sup 3/He reactions play the main role. It also discusses possible applications of these small generators as thermalneutron sources, addresses the small accelerators as charged particle and X-ray sources, enables suitable topics to be selected for education and training and provides a wide range of experiments with the detection of neutrons and charged particles, including the study of shielding and the generator technology itself.

The model of three-body Borromean halo nuclei breakup was described by using standard phase space distributions and the Monte Carlo simulation method was established to resolve the detection problem of two neutrons produced from breakup reaction on the neutron wall detector. For $^{6}$He case, overall resolution $\\sigma_{E_{k}}$ for the Gaussian part of the detector response and the detection efficiency including solid angle acceptance with regard to the excitation energy $E_{k}$ are obtained by the simulation of two neutrons from $^{6}$He breakup into the neutron wall. The effects of the algorithm on the angular and energy correlations of the fragments are briefly discussed.

We report radiation hardness tests performed at the Frascati Neutron Generator on silicon Photo-Multipliers, semiconductor photon detectors built from a square matrix of avalanche photo-diodes on a silicon substrate. Several samples from different manufacturers have been irradiated integrating up to 7x10^10 1-MeV-equivalent neutrons per cm^2. Detector performances have been recorded during the neutron irradiation and a gradual deterioration of their properties was found to happen already after an integrated fluence of the order of 10^8 1-MeV-equivalent neutrons per cm^2.

We report radiation hardness tests performed at the Frascati Neutron Generator on silicon Photo-Multipliers, semiconductor photon detectors built from a square matrix of avalanche photo-diodes on a silicon substrate. Several samples from different manufacturers have been irradiated integrating up to 7x10^10 1-MeV-equivalent neutrons per cm^2. Detector performances have been recorded during the neutron irradiation and a gradual deterioration of their properties was found to happen already after an integrated fluence of the order of 10^8 1-MeV-equivalent neutrons per cm^2.

Measurements were performed with a single annular, stainless-steel-canned casting of uranium (93.17 wt% 235U) metal ( ~18 kg) to provide data to verify calculational methods for criticality safety. The measurements used a small portable DT generator with an embedded alpha detector to time and directionally tag the neutrons from the generator. The center of the time and directional tagged neutron beam was perpendicular to the axis of the casting. The radiation detectors were 1x1x6 in plastic scintillators encased in 0.635-cm-thick lead shields that were sensitive to neutrons above 1 MeV in energy. The detector lead shields were adjacent to the casting and the target spot of the generator was about 3.8 cm from the casting at the vertical center. The time distribution of the fission induced radiation was measured with respect to the source event by a fast (1GHz) processor. The measurements described in this paper also include time correlation measurements with a time tagged spontaneously fissioning 252Cf neutron source, both on the axis and on the surface of the casting. Measurements with both types of sources are compared. Measurements with the DT generator closely coupled with the HEU provide no more additional information than those with the Cf source closely coupled with the HEU and are complicated by the time and directionally tagged neutrons from the generator scattering between the walls and floor of the measurements room and the casting while still above detection thresholds.

The development of a three-dimensional coupled neutronics/thermalhydraulics code for LWR safety analysis has been initiated. The transient neutronics code QUANDRY has been joined to the two-fluid thermal-hydraulics code ...

Tritium has been detected evolving from samples of deuteriated palladium wires and powders subject to pulsed high voltage at Los Alamos. They wanted to measure whether these samples were emitting neutrons. The idea of pulsing current through the wires and powders was to drive the deuterium in and out by rapid electrical heating. With promising tritium results in hand, the experiments were prepared at Los Alamos, and then taken to BYU and run in the neutrondetector located in a tunnel in Provo canyon under 35 m of rock and dirt overburden. The neutronsdetector and sample setup are described. Results including total neutron counts, time distributions, and an indication of the energy distributions are discussed. The results do not provide compelling evidence of neutron production, but are not inconsistent with earlier measurements of neutrons and tritium. Difficulties in explaining the difference in tritium and neutron measurements are also discussed. Plans for further work are presented.

Comprehensive treatment of neutron interactions in condensed matter at energies from thermal to MeV, focusing on aspects most relevant to radiation therapy, industrial imaging, and materials research applications. Comparative ...

US/Japan Wide-Angle Neutron Diffractometer US/Japan Wide-Angle Neutron Diffractometer WAND Instrument scientist Jaime Fernandez-Baca (left) with a visiting researcher at WAND. The Wide-Angle Neutron Diffractometer (WAND) at the HFIR HB-2C beam tube was designed to provide two specialized data-collection capabilities: (1) fast measurements of medium-resolution powder-diffraction patterns and (2) measurements of diffuse scattering in single crystals using flat-cone geometry. For these purposes, this instrument is equipped with a curved, one-dimensional 3He position-sensitive detector covering 125Âș of the scattering angle with the focal distance of 71 cm. The sample and detector can be tilted in the flat-cone geometry mode. These features enable measurement of single-crystal diffraction patterns in a short time over a

To meet the needs for neutron capture theory (NCT) irradiations, a high-intensity, high-quality fusion converter-based epithermal neutron beam has been designed for the MITR-II research reactor. This epithermal neutron beam, capable of delivering treatments in a few minutes with negligible background contamination from fast neutrons and photons, will be installed in the present thermal column and hohlraum of the 5-MW MITR-II research reactor. Spent or fresh MITR-II fuel elements will be used to fuel the converter. With a fission converter power of {approximately}80 kW using spent fuel, epithermal fluxes (1 eV < E < 10 keV) in excess of 10{sup 10} n/cm{sup 2} {center_dot} s are achievable at the target position with negligible photon and fast neutron contamination, i.e., <2 {times} 10{sup {minus}11}cGy-cm{sup 2}/n. With the currently available {sup 10}B delivery compound boronophenylalanine-fructose, average therapeutic ratios of {approximately}5 can be achieved using this beam for brain irradiations with deep effective penetration ({approximately}9.5 cm) and high dose rates of up to 400 to 600 RBE cGy/min. If NCT becomes an accepted therapy, fission converter-based beams constructed at existing reactors could meet a large fraction of the projected requirements for intense, low-background epithermal neutron beams at a relatively low cost. The results of an extensive set of neutronic design studies investigating all components of the beam are presented. These detailed studies can be useful as guidance for others who may wish to use the fission converter approach to develop epithermal beams for NCT.

A fast neutron nuclear reactor including a core and a plurality of vertically oriented neutron shield assemblies surrounding the core. Each assembly includes closely packed cylindrical rods within a polygonal metallic duct. The shield assemblies are less susceptible to thermal stresses and are less massive than solid shield assemblies, and are cooled by liquid coolant flow through interstices among the rods and duct.

Failure detectors -- oracles that provide information about process crashes -- are an important abstraction for crash tolerance in distributed systems. The generality of failure-detector theory, while providing great ...

Cold Neutron Chopper Spectrometer at SNS Cold Neutron Chopper Spectrometer at SNS CNCS detector array Interior of the CNCS detector array. CNCS is a high-resolution, direct-geometry, multi-chopper inelastic spectrometer designed to provide flexibility in the choice of energy resolution and to perform best at low incident energies (2 to 50 meV). Although the detector coverage around the sample is 1.7 sr, a later upgrade to 3 sr is possible. Experiments at CNCS typically use energy resolutions between 10 and 500 Â”eV. A broad variety of scientific problems, ranging from complex and quantum fluids to magnetism and chemical spectroscopy, are being addressed through experiments at CNCS. Applications Schematic of CNCS (larger image). Complex fluids: dilute protein solutions, biological gels, selective

A gamma ray detector shield comprised of a rigid, lead, cylindrical-shaped vessel having upper and lower portions with an pneumatically driven, sliding top assembly. Disposed inside the lead shield is a gamma ray scintillation crystal detector. Access to the gamma detector is through the sliding top assembly.

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Local features are the building blocks of many visual systems, and local point detector is usually the first component for local feature extraction. Existing local point detector are designed with target for matching and it may not perform well when ... Keywords: semantic point detector

OF LOW-ENERGY NEUTRONS IN SOLAR FLARES AND THE IMPORTANCE OF THEIR DETECTION IN THE INNER HELIOSPHERE R 20375, USA; murphy@ssd5.nrl.navy.mil 2 Department of Physics and Astronomy, Tel Aviv University, Tel ABSTRACT Neutrondetectors on spacecraft in the inner heliosphere can observe the low-energy (

A neutron source of the antimony--beryllium type is presented. The source is comprised of a solid mass of beryllium having a cylindrical recess extending therein and a cylinder containing antimony-124 slidably disposed within the cylindrical recess. The antimony cylinder is encased in aluminum. A berylliunn plug is removably inserted in the open end of the cylindrical recess to completely enclose the antimony cylinder in bsryllium. The plug and antimony cylinder are each provided with a stud on their upper ends to facilitate handling remotely.

A neutron-detection system was built for the purpose of measuring the neutron flux from an Inertial-Electrostatic Confinement Device located at Brigham Young University. A BF$sub 3$ proportional counter was used for absolute flux measurements and a pair of scintillation detectors was used to compare neutron output under different operating conditions. The detectors were designed to be compatible with the operating conditions of the device and to be able to measure small changes in neutron output. The detectors were calibrated using a Pu-Be source with corrections made for laboratory conditions. Performance of the counting system was checked and data were collected on the neutron flux from the device. (auth)

Noble gas excimer detectors are a technology that is common in particle physics research and less common in applications for security and international safeguards. These detectors offer the capability to detect gammas with an energy resolution similar to NaI and to detect neutrons with good energy resolution as well. Depending on the noble gas selected and whether or not it is in a gaseous or liquid state, the sensitivity to gammas and neutrons can be tuned according to the needs of the application. All of this flexibility can be available at a significant cost saving over alternative technologies. This paper will review this detector technology and its applicability to security and safeguards.

The preferred embodiments are directed to a high-energy detector that is electrically shielded using an anode, a cathode, and a conducting shield to substantially reduce or eliminate electrically unshielded area. The anode and the cathode are disposed at opposite ends of the detector and the conducting shield substantially surrounds at least a portion of the longitudinal surface of the detector. The conducting shield extends longitudinally to the anode end of the detector and substantially surrounds at least a portion of the detector. Signals read from one or more of the anode, cathode, and conducting shield can be used to determine the number of electrons that are liberated as a result of high-energy particles impinge on the detector. A correction technique can be implemented to correct for liberated electron that become trapped to improve the energy resolution of the high-energy detectors disclosed herein.

BABAR Detector BABAR Detector This page provides background information on HEP detectors in general and the BaBar detector in particular. A Technical Introduction to Particle-Physics Experiments A particle physics experiment has two basic components: an accelerator and a detector. The particle accelerator's job is to produce the high-energy particles. It does this by taking a particle, speeding it up using electromagnetic fields, and crashing it into another particle. At first, only one or two high-energy particles are produced, but these soon decay to many more lower-energy particles, so you end up with lots of particles shooting out from the collision point. The detector's job is to record information about the particles. A typical particle detector consists of several subdetectors, each of which performs a different type of measurement. Particles from the collision pass through and interact with each subdetector, and the results are recorded.

We present the design of an instrument capable of measuring the high energy ($>$60 MeV) muon-induced neutron flux deep underground. The instrument is based on applying the Gd-loaded liquid-scintillator technique to measure the rate of high-energy neutrons underground based on the neutron multiplicity induced in a Pb target. We present design studies based on Monte Carlo simulations that show that an apparatus consisting of a Pb target of 200 cm by 200 cm area by 60 cm thickness covered by a 60 cm thick Gd-loaded liquid scintillator (0.5% Gd content) detector could measure, at a depth of 2000 meters of water equivalent, a rate of $70\\pm8$ (stat) events/year. Based on these studies, we also discuss the benefits of using a neutron multiplicity meter as a component of active shielding in such experiments.

We present a new approach to measuring the neutrino-spin correlation parameter B in neutron decay. The approach combines the technology of large-area ion-implanted silicon detectors being developed for the abBA experiment, with an ultracold neutron source to provide more precise neutron polarimetry. The technique detects both proton and electron from the neutron decay in coincidence. B is determined from an electron-energy-dependent measurement of the proton spin asymmetry. This approach will provide a statistical precision of 1 x 10-4 . The systematic precision is still being evaluated, but is expected to be below 1 x 10-3 , and could approach 1 x 10-4 . A measurement of B with this precision would place constraints on supersymmetric extensions to the Standard Model.

A neutron flux monitor is provided which consists of a plurality of fission counters arranged as spaced-apart point detectors along a delay line. As a fission event occurs in any one of the counters, two delayed current pulses are generated at the output of the delay line. The time separation of the pulses identifies the counter in which the particular fission event occured. Neutron flux profiles of reactor cores can be more accurately measured as a result.

Monte Carlo simulations of simple neutron oil well logging tools into typical geological formations are presented.The simulated tools consist of both 14 MeV pulsed and continuous Am-Be neutron sources with time gated and continuous gamma ray detectors respectively.The geological formation consists of pure limestone with 15% absolute porosity in a wide range of oil saturation.The particle transport was performed with the Monte Carlo N-Particle Transport Code System, MCNP-4B.Several gamma ray spectra were obtained at the detector position that allow to perform composition analysis of the formation.In particular, the ratio C/O was analyzed as an indicator of oil saturation.Further calculations are proposed to simulate actual detector responses in order to contribute to understand the relation between the detector response with the formation composition

The ArDM project aims at operating a large noble liquid detector to search for direct evidence of Weakly Interacting Massive Particles (WIMP) as Dark Matter in the universe. Background sources relevant to ton-scale liquid and gaseous argon detectors, such as neutrons from detector components, muon-induced neutrons and neutrons caused by radioactivity of rock, as well as the internal $^{39}Ar$ background, are studied with simulations. These background radiations are addressed with the design of an appropriate shielding as well as with different background rejection potentialities. Among them the project relies on event topology recognition, event localization, density ionization discrimination and pulse shape discrimination. Background rates, energy spectra, characteristics of the background-induced nuclear recoils in liquid argon, as well as the shielding performance and rejection performance of the detector are described.

The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) resumed full power operations on May 16, 2007. There were three experiment cycles of 23 to 25 days in FY2007 and another six are proposed for FY2008 beginning in November 2007. During FY 2007, the High Flux Isotope Reactor delivered 1178 operating hours to users. Commissioning of two SANS instruments is under way and these instruments will join the user program in 2008. The Neutron Scattering Science Advisory Committee endorsed language encouraging development of the science case for two instruments proposed for HFIR.

The Yankee nuclear power station located in Rowe, Massachusetts, permanently ceased power operations on February 26, 1992, after 31 yr of operation. Yankee has since initiated decommissioning planning activities. A significant component of these activities is a determination of the extent of radiological contamination of the Yankee site. Included in this effort was determination of the extent of neutron activation of plant components. This paper describes the determination of the neutron activation of the Yankee reactor vessel, associated internals, and surrounding structures. The Yankee reactor vessel is a 600-MW(thermal) stainless steel-lined, carbon steel vessel with stainless steel internal components designed by Westinghouse. The reactor vessel is surrounded and supported by a carbon steel neutron shield tank that was filled with chromated water during plant operation. A 5-ft-thick concrete biological shield wall surrounds the neutron shield tank. A project is under way to remove the reactor vessel internals from the reactor vessel.

In a PHYSOR 2010 paper(1) we introduced a fusion hybrid reactor whose fusion component is the gasdynamic mirror (GDM), and whose blanket was made of thorium - 232. The thrust of that study was to demonstrate the performance of such a reactor by establishing the breeding of uranium - 233 in the blanket, and the burning thereof to produce power. In that analysis, we utilized the diffusion equation for one-energy neutron group, namely, those produced by the fusion reactions, to establish the power distribution and density in the system. Those results should be viewed as a first approximation since the high energy neutrons are not effective in inducing fission, but contribute primarily to the production of actinides. In the presence of a coolant, however, such as water, these neutrons tend to thermalize rather quickly, hence a better assessment of the reactor performance would require at least a two group analysis, namely the fast and thermal groups. We follow that approach and write an approximate set of equations for the fluxes of these groups. From these relations we deduce the all-important quantity, k{sub eff}, which we utilize to compute the multiplication factor, and subsequently, the power density in the reactor. We show that k{sub eff} can be made to have a value of 0.99, thus indicating that 100 thermalneutrons are generated per fusion neutron, while allowing the system to function as 'subcritical.' Moreover, we show that such a hybrid reactor can generate hundreds of megawatts of thermal power per cm of length depending on the flux of the fusion neutrons impinging on the blanket. (authors)

The cooling and reheating histories of dim isolated neutron stars(DINs) are discussed. Energy dissipation due to dipole spindown with ordinary and magnetar fields, and due to torques from a fallback disk are considered as alternative sources of reheating which would set the temperature of the neutron star after the initial cooling era. Cooling or thermal ages are related to the numbers and formation rates of the DINs and therefore to their relations with other isolated neutron star populations. Interaction with a fallback disk, higher multipole fields and activity of the neutron star are briefly discussed.

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The quality of a neutron-imaging beam directly impacts the quality of radiographic images produced using that beam. Fully characterizing a neutron beam, including determination of the beam's effective length-to-diameter ratio, neutron flux profile, energy spectrum, potential image quality, and beam divergence, is vital for producing quality radiographic images. This paper provides a characterization of the east neutron imaging beamline at the Idaho National Laboratory Neutron Radiography Reactor (NRAD). The experiments which measured the beam's effective length-to-diameter ratio and potential image quality are based on American Society for Testing and Materials (ASTM) standards. An analysis of the image produced by a calibrated phantom measured the beam divergence. The energy spectrum measurements consist of a series of foil irradiations using a selection of activation foils, compared to the results produced by a Monte Carlo n-Particle (MCNP) model of the beamline. The NRAD has an effective collimation ratio greater than 125, a beam divergence of 0.3 +_ 0.1 degrees, and a gold foil cadmium ratio of 2.7. The flux profile has been quantified and the facility is an ASTM Category 1 radiographic facility. Based on bare and cadmium covered foil activation results, the neutron energy spectrum used in the current MCNP model of the radiography beamline over-samples the thermal region of the neutron energy spectrum.

The distribution of neutrons with energies below 15 MeV in spherical stony meteoroids is calculated using the ANISN neutron-transport code. The source distributions and intensities of neutrons are calculated using cross sections for the production of tritium. The meteoroid's radius and chemical composition strongly influence the total neutron flux and the neutron energy spectrum, while the location within a meteoroid only affects the relative neutron intensities. Meteoroids need to have radii of more than 50 g/cm/sup 2/ before they have appreciable fluxes of neutrons near thermal energies. Meteoroids with high hydrogen or low iron contents can thermalizeneutrons better than chondrites. Rates for the production of /sup 60/Co, /sup 59/Ni, and /sup 36/Cl are calculated with evaluated neutron-capture cross sections and neutron fluxes determined for carbonaceous chondrites with high hydrogen contents, L-chondrites, and aubrites. For most meteoroids with radii < 300 g/cm/sup 2/, the production rates of these neutron-capture nuclides increase monotonically with depth. The highest calculated /sup 60/Co production rate in an ordinary chondrite is 375 atoms/(min g-Co) at the center of a meteoroid with a 250 g/cm/sup 2/ radius. The production rates calculated for spallogenic /sup 60/Co and /sup 59/Ni are greater than the neutron-capture rates for radii less than approx.50-75 g/cm/sup 2/. Only for very large meteoroids and chlorine-rich samples is the neutron-capture production of /sup 36/Cl important. The results of these calculations are compared with those of previous calculations and with measured activities in many meteorites. 44 refs., 15 figs., 1 tab.

A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

An improved neutron reflecting supermirror structure comprising a plurality of stacked sets of bilayers of neutron reflecting materials. The improved neutron reflecting supermirror structure is adapted to provide extremely good performance at high incidence angles, i.e. up to four time the critical angle of standard neutron mirror structures. The reflection of neutrons striking the supermirror structure at a high critical angle provides enhanced neutron throughput, and hence more efficient and economical use of neutron sources. One layer of each set of bilayers consist of titanium, and the second layer of each set of bilayers consist of an alloy of nickel with carbon interstitially present in the nickel alloy.

A new type of solar neutrondetector (SEDA-FIB) was launched on board the Space Shuttle Endeavor on July 16 2009, and began collecting data at the International Space Station (ISS) on August 25 2009. This paper summarizes four years of observations with the solar neutrondetector SEDA-FIB (Space Environment Data Acquisition using the FIBer detector). The solar neutrondetector FIB can determine both the energy and arrival direction of solar neutrons. In this paper, we first present the angular distribution of neutron induced protons obtained in Monte Carlo simulations. The results are compared with the experimental results. Then we provide the angular distribution of background neutrons during one full orbit of the ISS (90 minutes). Next, the angular distribution of neutrons during the flare onset time from 20:02 to 20:10 UT on March 7 2011 is presented. It is compared with the distribution when a solar flare is not occurring. Observed solar neutrons possibly originated from the M-class solar flares that occu...

The Weapons Neutron Research Facility has been used to obtain moderate-resolution total neutron cross section data for H, C, /sup 208/Pb, /sup 232/Th, /sup 238/U, and /sup 242/Pu over the energy range 5 to 200 MeV. Neutrons were produced by bombarding a 2.5-cm diam by 15-cm long Ta target with an 800 MeV pulsed proton beam from LAMPF. A 10.2-cm diam by 15.2-cm thick NE110 proton recoil detector was used at a flight path of 32 meters, giving a time-of-flight resolution of 60 ps/m. The total cross section results are compared to ENDF/BV evaluations and to previous data where possible.

We describe a novel experimental technique for neutron imaging with scattered neutrons. These scattered neutrons are of interest for condensed matter physics, because they permit to reveal the local distribution of incoherent and coherent scattering within a sample. In contrast to standard attenuation based imaging, scattered neutron imaging distinguishes between the scattering cross section and the total attenuation cross section including absorption. First successful low-noise millimeter-resolution images by scattered neutron radiography and tomography are presented.

High quality radioactive beams have recently made possible the measurement of (d,p) reactions on unstable nuclei in inverse kinematics, which can yield information on the development of single-neutron structure away from stability, and are of astrophysical interest due to the proximity to suggested r-process paths. The Oak Ridge Rutgers University Barrel Array (ORRUBA) is a new high solid-angular coverage array, composed of two rings of silicon detectors, optimized for measuring (d,p) reactions. A partial implementation has been used to measure (d,p) reactions on nuclei around the N = 82 shell closure.

The smuggling of special nuclear material (SNM) has long been a concern. In April 2009, President Obama declared that a terrorist acquiring a nuclear weapon was the most immediate threat to global security. The Second Line of Defense (SLD) initiative was stood up by the National Nuclear Security Administration to deter, detect, and interdict illicit trafficking of nuclear and radioactive materials across international borders and maritime shipping. The SLD initiative does not provide for the detection of SNM being carried on small, personal watercraft. Previous work examined the possibility of using active neutrondetectors to induce fission in SNM and detect the response. This thesis examines the possibility of detecting SNM using passive 3He neutrondetectors. Monte Carlo N-Particle (MCNP) simulations were run to determine the best detector configuration. Detecting sources at increasing depths, detecting moving sources and the effects of waves were also simulated in MCNP. Comparisons with experimental measurements showed that detectors parallel to the surface of water were best at detecting neutron sources below the surface. Additionally, stacking detectors and placing a cadmium sheet between the polyethylene blocks resulted in a greater ability to determine the height of a source by taking the ratio of count rates in the lower and upper detectors. Using this configuration, a source of strength 3.39 x 10^5 n/s could be detected to a depth of 12.00 in below the water surface. Count rates in the presence of waves did not average out to count rates taken above a flat plane of water. Detectors closer to the water performed worse than above a flat plane while detectors placed higher recorded more counts than above a flat plane. Moving sources were also simulated; sources under water, 3.00 ft from the detectors, and moving at 5.8 kts could be detected above background.

This document serves as both an FY2103 End-of-Year and End-of-Project report on efforts that resulted in the design of a prototype fast neutron multiplicity counter leveraged upon the findings of previous project efforts. The prototype design includes 32 liquid scintillator detectors with cubic volumes 7.62 cm in dimension configured into 4 stacked rings of 8 detectors. Detector signal collection for the system is handled with a pair of Struck Innovative Systeme 16-channel digitizers controlled by in-house developed software with built-in multiplicity analysis algorithms. Initial testing and familiarization of the currently obtained prototype components is underway, however full prototype construction is required for further optimization. Monte Carlo models of the prototype system were performed to estimate die-away and efficiency values. Analysis of these models resulted in the development of a software package capable of determining the effects of nearest-neighbor rejection methods for elimination of detector cross talk. A parameter study was performed using previously developed analytical methods for the estimation of assay mass variance for use as a figure-of-merit for system performance. A software package was developed to automate these calculations and ensure accuracy. The results of the parameter study show that the prototype fast neutron multiplicity counter design is very nearly optimized under the restraints of the parameter space.

Fissile materials emit neutrons with an unmistakable signature that can reveal characteristics of the material. We describe here measurements, simulations, and predicted signals expected and prospects for application of neutron correlation measurement methods to detection of special nuclear materials (SNM). The occurrence of fission chains in SNM can give rise to this distinctive, measurable time correlation signal. The neutron signals can be analyzed to detect the presence and to infer attributes of the SNM and surrounding materials. For instance, it is possible to infer attributes of an assembly containing a few kilograms of uranium, purely passively, using detectors of modest size in a reasonable time. Neutron signals of three radioactive sources are shown to illustrate the neutron correlation and analysis method. Measurements are compared with Monte Carlo calculations of the authenticated sources.

A neutron imaging diagnostic has recently been commissioned at the National Ignition Facility (NIF). This new system is an important diagnostic tool for inertial fusion studies at the NIF for measuring the size and shape of the burning DT plasma during the ignition stage of Inertial Confinement Fusion (ICF) implosions. The imaging technique utilizes a pinhole neutron aperture, placed between the neutron source and a neutrondetector. The detection system measures the two dimensional distribution of neutrons passing through the pinhole. This diagnostic has been designed to collect two images at two times. The long flight path for this diagnostic, 28 m, results in a chromatic separation of the neutrons, allowing the independently timed images to measure the source distribution for two neutron energies. Typically the first image measures the distribution of the 14 MeV neutrons and the second image of the 6-12 MeV neutrons. The combination of these two images has provided data on the size and shape of the burning plasma within the compressed capsule, as well as a measure of the quantity and spatial distribution of the cold fuel surrounding this core.

Abstract Cherenkov detectors are widely used for particle identification and threshold detectors in high-energy physics. Glass Cherenkov detectors that are sensitive to beta emissions originating from neutron activation have been demonstrated recently as a potential replacement for activation foils. In this work, we set the groundwork to evaluate large Cherenkov glass detectors for sensitivity to MeV photons through first understanding the measured response of small Cherenkov glass detectors to isotopic gamma-ray sources. Counting and pulse height measurements are acquired with reflected glass Cherenkov detectors read out with a photomultiplier tube. Simulation was used to inform our understanding of the measured results. This simulation included radioactive source decay, radiation interaction, Cherenkov light generation, optical ray tracing, and photoelectron production. Implications for the use of Cherenkov glass detectors to measure low energy gammaray response are discussed.

Sample records for thermal neutron detector from the National Library of Energy Beta (NLEBeta)

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